Federal Register / Vol. 63, No. 106 /Wednesday, June 3, 1998 / Notices 30261

MATTERS TO BE CONSIDERED:

Week of June 1

Tuesday, June 2

8:00 a.m.—Briefing on Remaining Issues

Related to Proposed Restart of

Millstone Unit 3. (Public Meeting)

(Contact: Bill Travers, 301–415–

1200).

1:00 p.m.—(Continuation of morning

meeting on Millstone.)

Wednesday, June 3

2:00 p.m.—Briefing by the Executive

Branch (Closed—Ex. 1).

Thursday, June 4

2:00 p.m.—Briefing by NEI and NRC

Staff on Safety Evaluations, FSAR

Updates and Incorporation of Risk

Insights (Public Meeting).

Friday, June 5

10:00 a.m.—Briefing by EPRI on the

Status of their Advanced Light

Water Reactor (ALWR) Program

(Public Meeting).

11:30 a.m.—Affirmation Session (Public

Meeting); a: Private Fuel Storage,

L.L.C.; Ruling by Chief Judge of the

Atomic Safety and Licensing Board

Panel to Establish a Second Board,

LBP–98–8 (April 24, 1998).

Week of June 8—Tentative

Thursday, June 11

11:30 a.m.—Affirmation Session (Public

Meeting) (if needed).

Friday, June 12

10:00 a.m.—Briefing by Reactor Vendors

Owners’ Groups (Public Meeting)

(Contact: Bryan Sheron, 301–415–

1274).

Week of June 15—Tentative

Wednesday, June 17

10:00 a.m.—Briefing by National Mining

Association on Regulation of the

Uranium Recovery Industry (Public

Meeting).

11:30 a.m.—Affirmation Session (Public

Meeting) (if needed).

2:00 p.m.—Meeting with Advisory

Committee on Medical Uses of

Isotopes (ACMUI) and Briefing on

Part 35 QM Rule (Public Meeting)

(Contact: Larry Camper, 301–415–

7231).

Week of June 22—Tentative

Thursday, June 25

9:30 a.m.—Briefing by IG on Results of

NRC Organization Safety Culture

and Climate Survey (Public

Meeting).

11:30 a.m.—Affirmation Session (Public

Meeting) (if needed).

2:00 p.m.—Briefing on EEO Program

(Public Meeting).

*The schedule for Commission meetings is

subject to change on short notice. To verify

the status of meetings call (recording)—(301)

415–1292. Contact person for more

information: Bill Hill (301) 415–1661.

The NRC Commission Meeting

Schedule can be found on the Internet

at:

http://www.nrc.gov/SECY/smj/

schedule/htm

This notice is distributed by mail to

several hundred subscribers; if you no

longer wish to receive it, or would like

to be added to it, please contact the

Office of the Secretary, Attn: Operations

Branch, Washington, D.C. 20555 (301–

415–1661). In addition, distribution of

this meeting notice over the Internet

system is available. If you are interested

in receiving this Commission meeting

schedule electronically, please send an

electronic message to wmh@nrc.gov or

dkw@nrc.gov.

Dated: May 29, 1998.

William M. Hill, Jr.,

Secy, Tracking Officer, Office of the Secretary.

[FR Doc. 98–14821 Filed 6–1–98; 11:15 am]

BILLING CODE 7590–01–M

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97–415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from May 11,

1998, through May 21, 1998. The last

biweekly notice was published on May

20, 1998 (63 FR 27757).

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
30262 Federal Register / Vol. 63, No. 106 /Wednesday, June 3, 1998 / Notices

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC.

The filing of requests for a hearing and

petitions for leave to intervene is

discussed below.

By July 6, 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

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: March 6,

1998.

Description of amendment request:

The proposed change will revise the H.

B. Robinson, Unit 2 Technical

Specifications to allow use of the Post

Accident Monitoring (PAM) source

range (SR) neutron flux detector as a

compensatory measure in the event that

one of the two required BF3 detectors

become inoperable while the plant is in

MODE 6.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The proposed change to Technical

Specifications is only applicable during

the refueling mode of operation (MODE

6). Neither the BF3 SR nor PAM neutron

flux monitors provide an automatic

initiation signal for the operation of

plant systems or components but are

only relied upon to provide indication

of core reactivity. Since the proposed

change to Technical Specifications does

not alter the design or operation of plant

equipment or systems, there is no

change in the initiating mechanisms for
Federal Register / Vol. 63, No. 106 /Wednesday, June 3, 1998 / Notices 30263

any accidents previously analyzed.

Therefore this change does not involve

a significant increase in the probability

for an accident previously analyzed.

The UFSAR [Updated Final Safety

Analysis Report] identifies two

accidents that credit the SR monitoring

capability in MODE 6, the boron

dilution accident and the fuel handling

accident. No other accidents were found

to rely on SR monitoring in MODE 6.

The proposed change will continue to

require BF3 SR visual indication of core

reactivity in the control room and a BF3

SR neutron flux monitor audible

indication in containment. This change

will not result in a significant reduction

in operator capability to detect

unexpected changes in core reactivity

and perform actions credited with

termination of those events, therefore

the proposed change does not involve a

significant increase in the consequences

of an accident previously analyzed.

2. Does the change create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

The proposed change to Technical

Specifications does not involve any

physical alteration of plant systems,

structures or components or changes in

parameters governing plant operations.

The proposed change will not result in

a significant reduction in monitoring

capability since two BF3 SR channels of

SR visual indication in the control room

and audible SR indication in the

containment are required during core

alterations and positive reactivity

changes. The use of the PAM SR

neutron flux monitor as a compensatory

measure does not introduce any new

accident initiation scenarios since the

SR instruments are for monitoring and

criticality assessment only and are not

relied upon to initiate automatic

accident mitigation measures.

Therefore, this change does not create

the possibility of a new or different kind

of accident from any accident

previously analyzed.

3. Does this change involve a

significant reduction in a margin of

safety?

The proposed change will maintain

two BF3 SR monitoring means for

visually monitoring core reactivity as

currently discussed in the bases for the

affected Technical Specifications.

Audible indication provided by one BF3

SR neutron flux monitor will still be

required and fulfilled by the remaining

BF3 SR neutron flux monitor. The PAM

SR neutron flux monitors use fission

chambers as detectors which have a

sensitivity of 4 cps/neutron-volts (cps/

nv) for thermal neutrons and 2 cps/nv

for fast neutrons. The BF3 SR neutron

flux monitors have a sensitivity of 9

cps/nv. The PAM SR neutron flux

monitor has comparable range and

accuracy (i.e., range of 1E–01 cps to

1E+05 cps with an accuracy of 2% of

full scale) to that of BF3 SR neutron flux

monitor (i. e., range of 1E–00 cps to 1E

+ 06 cps with an accuracy of 3% of full

scale) which meets the Technical

Specifications Section 3.9.2 Bases

requirements of 6 decades of indication

and 5% accuracy. Therefore, this change

does not involve a significant reduction

in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: 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: P. T. Kuo,

Acting.

Commonwealth Edison Company,

Docket No. 50–249, Dresden Nuclear

Power Station, Unit 3, Grundy County,

Illinois

Date of amendment request: May 6,

1998.

Description of amendment request:

The proposed amendment would amend

Technical Specification (TS) 4.6.E to

allow a one-time extension of the 40-

month requirement to pressure set test

or replace all Main Steam Safety Valves

(MSSVs) to a maximum interval of 60

months as currently allowed by the

American Society of Mechanical

Engineers (ASME) Boiler and Pressure

Vessel Code (Code).

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 because

of the following:

The proposed changes request a onetime change to the surveillance

requirement for the MSSVs. The

surveillance interval between safety

valve testing is not a precursor assumed

in any previously analyzed accident.

Therefore, the probability of a

previously evaluated accident has not

been increased.

The proposed extension is consistent

with the ASME Code requirement to test

all valves within 60 months. The

proposed changes are also consistent

with NUREG–1433 and do not adversely

affect existing plant safety margins or

the reliability of the equipment assumed

to operate in the safety analysis.

Operating experience and superior

materiel condition of the MSSVs

support the expectation that they will

continue to perform their intended

function. Therefore, the consequences of

a previously evaluated accident have

not been increased.

2. Create the possibility of a new or

different kind of accident from any

accident previously evaluated because:

No new equipment is required, nor

will the MSSVs be operated in a

different manner during the period of

the extended surveillance interval. The

proposed change is consistent with

NUREG–1433 requirements for safety

valve surveillance intervals as well as

the ASME Code for requirements testing

safety valves. Operating experience and

superior materiel condition of the

MSSVs support the expectation that

they will continue to perform their

intended function. Therefore, the

possibility of a new or different accident

has not been increased.

3. Involve a significant reduction in

the margin of safety because:

The proposed amendment represents

an extension to the current TS

requirements, but would otherwise be

provided generically by ASME Code.

The proposed changes are also

consistent with NUREG–1433, request a

shorter total interval than previously

granted by the Staff (Reference b), [J.F.

Stang (NRC) to D. L. Farrar, SER dated

October 8, 1996] and do not adversely

affect existing plant safety margins or

the reliability of the equipment assumed

to operate in the safety analysis. The

proposed changes have been evaluated

and found to be acceptable for use at

Dresden based on system safety analysis

requirements and operational

performance. The MSSV provisions

continue to be adequately maintained

during plant operation. The proposed

changes to the MSSV surveillance

interval do not significantly reduce

existing plant safety margins since

excellent materiel condition and

acceptable surveillance test results

support the expectation that no

significant degradation will occur over

the extended interval.

The proposed changes are based on

NRC accepted provisions at other

operating plants that are applicable at
30264 Federal Register / Vol. 63, No. 106 /Wednesday, June 3, 1998 / Notices

Dresden and maintain necessary levels

of system or component reliability.

The proposed amendment for Dresden

will not reduce the availability of

systems required to mitigate accident

conditions; therefore, the proposed

changes do not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: Morris Area Public Library

District, 604 Liberty Street, Morris,

Illinois 60450.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603.

NRC Project Director: Stuart A.

Richards.

Illinois Power Company, Docket No. 50–

461, Clinton Power Station, Unit 1,

DeWitt County, Illinois

Date of amendment request: May 4,

1998.

Description of amendment request:

The proposed amendment would

incorporate Technical Specifications

Requirements for the protection systems

for the new static VAR compensators

being installed onsite to address

degraded electrical grid voltage.

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 changes addressed by this

amendment request involve the addition

of SVCs and their associated protection

systems to the onsite circuit connections

for the plant offsite electrical power

sources, i.e., to the RAT and ERAT. As

noted throughout this request, the

addition of the SVCs will help to

maintain voltage at the site for both of

the offsite electrical power sources

consistent with the ‘‘capacity and

capability’’ requirements of GDC 17.

Further, the regulating effect of the

SVCs will compensate for the voltage

drop that can occur without the SVCs

when the plant trips off-line (and thus

no longer supports grid voltage) during

normal or accident conditions. This

supports compliance with the GDC 17

requirement to minimize the probability

of losing electric power from the offsite

supplies as a result of, or coincident

with, the loss of power from the offsite

supplies as a result of, or coincident

with, the loss of power generated by the

nuclear power unit. Consequently, the

likelihood of transferring to the onsite

emergency power supplies (diesel

generators) during an accident will be

reduced. At the same time, as also

addressed in this amendment request,

incorporation of the SVCs into the CPS

auxiliary power system requires

consideration of failure modes that

could be introduced by the SVCs

wherein such failure modes could

involve a significant increase in the

probability or consequences of any

accident previously evaluated.

By supplying each of the SVCs with

an enhanced protection system,

consisting of dual, redundant protection

subsystems, either of which will isolate

the SVC from the bus (by automatically

opening the SVC main circuit breakers)

in response to postulated SVC failures

or associated abnormal conditions, the

potential for such conditions or failures

to adversely affect the plant safety

busses, the associated plant loads, or the

onsite emergency electrical power

sources is reduced to a very low

probability. The protection systems

designed for the SVCs include

consideration of failure modes or

abnormal conditions that may be

postulated or expected to occur with

some degree of probability for the offsite

electrical sources or grid with or

without the presence of the SVCs, (such

as a sustained degraded voltage

condition), as well as consideration of

any new or other failure modes or

abnormal conditions potentially

introduced by the SVCs that would be

less likely to occur in the offsite

electrical network without the presence

of the SVCs (such as the introduction of

harmonics). The proposed change to the

CPS Technical Specifications to

incorporate requirements for the SVC

protection systems will ensure that the

SVC protection systems are adequately

maintained in an operable condition to

perform their intended function of

protecting against such conditions or

failure modes. Operable SVC protection

systems will reduce the probability of

an SVC failure event that leads to

equipment damage and subsequent core

damage to a level that makes such an

event incredible.

It should be noted that tripping of the

SVCs in response to an SVC failure or

abnormal condition does not result in a

loss of power from the offsite sources.

Thus, the probability of a loss of offsite

power, which is an analyzed event in

the plant safety analyses, will not be

significantly increased by the SVC

protection systems.

As noted previously, the proposed

change to the Technical Specifications

to incorporate SVC protection system

operability and testing requirements

would ensure that plant safety systems

or components are not electrically

affected by the SVCs in an adverse

manner. In addition, except where the

SVCs are physically located and

connected to the ERAT and RAT via bus

ducts, plant safety-related structures

and supporting systems would not be

mechanically affected by the SVCs.

Separation, clearance and related

requirements to ensure no other

interaction with the RAT, ERAT and

offsite source connections, as well as for

maintaining offsite source

independence would be maintained. On

this basis, the safety functions of

systems for preventing or mitigating

analyzed events or accidents would not

be impacted by the SVCs.

Based on the above, the proposed

change to the Technical Specifications

does not involve a significant increase

in the probability or consequences of

any accident previously evaluated.

(2) In consideration of the potential

adverse impacts that the SVCs may have

on plant systems, structures or

components, such impacts are primarily

confined to potential electrical faults or

abnormal conditions. As noted above,

the SVCs have no mechanical impact on

safety-related plant systems, structures

or components. Thus, no new failure

modes or precursors to potentially new

and unanalyzed events would be

introduced via any mechanical means.

With respect to potential adverse

electrical impacts, the potential

electrical failure modes or abnormal

conditions postulated for the SVCs

include conditions or events that,

although could be considered possible

for the offsite sources (i.e., the grid),

were not in fact considered credible and

therefore previously evaluated for the

offsite electrical sources. These

conditions or events, such as the

introduction of harmonics or excessive

overvoltage or phase imbalance caused

by an SVC failure, would have the

potential to degrade plant safety-related

equipment connected to the busses at

the time of the SVC failure if no

protection for such conditions was

provided. However, enhanced

protection systems are provided for the

SVCs to ensure that such failures cannot

damage plant equipment. As noted

previously, the probability of an event

involving an SVC failure that leads to

equipment damage and subsequent core

damage has been calculated to be 1.5 x

10¥8/year. This low probability makes

such an event incredible just as

comparable events that could be

postulated for the offsite electrical

power sources were not previously
Federal Register / Vol. 63, No. 106 /Wednesday, June 3, 1998 / Notices 30265

considered credible and therefore were

not considered to be design basis events.

The calculated probability of 1.5 x

10¥8/year for an SVC failure event

involving core damage is an order of

magnitude lower than the threshold

probability criterion specified in Section

2.2.3 of the Standard Review Plan

(NUREG 0800) for design basis events

involving an offsite hazard that can lead

to core damage and radioactive release

with dose consequences in excess of the

limits specified in 10 CFR Part 100.

The proposed change to the Technical

Specifications incorporates

requirements for maintaining

operability of the SVC protection

systems. On this basis and as described

above, no new credible accidents that

could be associated with the SVCs (i.e.,

failure of the SVCs) are thus introduced,

so that the proposed change does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

(3) As noted previously, incorporation

of the SVCs into the CPS auxiliary

power system will support or regulate

plant bus voltage for both of the offsite

sources. Specifically, analysis has

shown that the SVCs will recover

reduced margin that has occurred or

would occur in the future (without the

SVCs) with respect to the voltage

required for plant safety loads and the

minimum expected offsite voltage,

under normal and accident conditions

(i.e., under steady-state and transient

voltage conditions). This also means

that the SVCs will enhance the

capability and capacity of the offsite

sources such that, when compared to

the configuration of not having the

SVCs, either source will be more likely

to reset the safety bus degraded voltage

relays in the event of an accident, thus

permitting the preferred offsite sources

to remain connected (and not causing a

transfer to the diesel generators). These

desirable results constitute a significant

increase in the margin of safety with

respect to voltage requirements for plant

loads.

Based on the above, IP has concluded

that the proposed change to the

Technical Specifications to support use

of the SVCs and their protection systems

does not involve a significant reduction

in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Vespasian Warner Public

Library, 120 West Johnson Street,

Clinton, IL 61727.

Attorney for licensee: Leah Manning

Stetzner, Vice President, General

Counsel, and Corporate Secretary, 500

South 27th Street, Decatur, IL 62525.

NRC Acting Project Director: Ronald

R. Bellamy.

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: April 22,

1998.

Description of amendment request:

The proposed change would revise

selected Technical Specification (TS)

surveillance requirements to

accommodate fuel cycles of up to 24

months for surveillances that are

currently performed at each 18-month

or other specified outage interval.

Specifically, the following TS

surveillance requirements would be

revised by the proposed change:

4.1.2.2.b and c, ‘‘Boration Systems Flow

Paths—Operating;’’ 4.3.3.5.2, ‘‘Remote

Shutdown System;’’ 4.4.3.2,

‘‘Pressurizer;’’ 4.4.4.1, ‘‘Relief Valves;’’

4.4.6.2.2.a and b, ‘‘Operational

Leakage;’’ 4.4.11.2, ‘‘Reactor Coolant

System Vents;’’ 4.5.1.1.d.1 and 2,

‘‘Accumulators;’’ 4.5.2.d, e, g.2), and h,

‘‘Emergency Core Cooling System

(ECCS) Subsystems—Tavg Greater Than

or Equal to 350°F;’’ 4.6.3.2,

‘‘Containment Isolation Valves;’’ and

4.7.1.2.1.c, ‘‘Auxiliary Feedwater

System.’’ In conjunction with the

proposed change, components

addressed in the following TS

surveillance requirements have been

evaluated to support an extension in

frequency to accommodate fuel cycles of

up to 24 months: 4.6.3.1 and 3,

‘‘Containment Isolation Valves;’’

4.7.1.2.2, ‘‘Auxiliary Feedwater

System;’’ 4.7.1.5, ‘‘Main Steam Line

Isolation Valves;’’ and 4.7.1.6,

‘‘Atmospheric Relief Valves.’’ In

addition, the proposed change would

delete the restriction ‘‘during

shutdown’’ in those TS surveillance

requirements where this restriction is

stated.

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 have no

adverse affect on accident initiators or

precursors nor alter the design

assumptions, conditions, configuration

of the facility or the manner in which

the plant is operated. The proposed

changes do not alter or prevent the

ability of structures, systems, or

components (SSCs) to perform their

intended function to mitigate the

consequences of an initiating event

within the acceptance limits assumed in

the Updated Final Safety Analysis

Report (UFSAR). The proposed changes

are administrative in nature and do not

change the level of programmatic

controls or the procedural details

associated with aforementioned

surveillance requirements.

Changing the frequencies of the

aforementioned surveillance

requirements from at least once per 18

months to at least once per refueling

interval does not change the basis for

the frequencies. The frequencies were

chosen because of the need to perform

these verifications under the conditions

that are normally found during a plant

refueling outage, and to avoid the

potential of an unplanned transient if

these surveillances were conducted

with the plant at power.

Equipment performance over several

operating cycles was evaluated to

determine the impact of extending the

surveillance intervals. This evaluation

included a review of surveillance

results, preventative maintenance

records, and the frequency and type of

corrective maintenance activities, and a

failure mode analysis. The evaluations

conclude that the subject SSCs are

highly reliable, presently exhibiting no

time dependent failure modes of

significance, and that there is no

indication that the proposed extension

could cause deterioration in the

condition or performance of the subject

SSCs. There are no known mechanisms

that would significantly degrade the

performance of the evaluated equipment

during normal plant operation.

Although there have been generic or

repetitive failures of some components

in the past, which may have affected the

ability of the SSCs to consistently and

successfully perform their safety

function, those items have been

resolved through design changes and

rework such that they have not recurred.

There have been no repetitive failures or

time dependent failures that were

significant in nature which would have

prevented the SSCs from performing

their intended safety function.

Deletion of the restriction ‘‘during

shutdown’’ where this restriction is

stated will permit performance of

certain maintenance and testing

activities during conditions or modes

other than shutdown. North Atlantic
30266 Federal Register / Vol. 63, No. 106 /Wednesday, June 3, 1998 / Notices

will ensure, through the implementation

of administrative controls that proper

regard to their effect on safe operation

of the plant is given prior to conduct of

a particular surveillance in a condition

or mode other than shutdown.

Since the proposed changes only

affect the surveillance intervals for SSCs

that are used to mitigate accidents, the

changes do not affect the probability or

consequence of a previously analyzed

accident. While the proposed changes

will lengthen the intervals between

surveillances, the increase in intervals

has been evaluated. Based on the

reviews of the surveillance tests,

inspections, and maintenance activities,

it is concluded that there is no

significant adverse impact on the

reliability or availability of these SSCs.

Since there are no changes to previous

accident analyses, the radiological

consequences associated with these

analyses remain unchanged, therefore,

the proposed changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed changes do not create

the possibility of a new or different kind

of accident from any previously

analyzed.

The proposed changes do not alter the

design assumptions, conditions,

configuration of the facility or the

manner in which the plant is operated.

There are no changes to the source term,

containment isolation or radiological

release assumptions used in evaluating

the radiological consequences in the

Seabrook Station UFSAR. Existing

system and component redundancy is

not being changed by the proposed

changes. The proposed changes have no

adverse impact on component or system

interactions. The proposed changes are

administrative in nature and do not

change the level of programmatic

controls and procedural details

associated with the aforementioned

surveillance requirements. Therefore,

since there are no changes to the design

assumptions, conditions, configuration

of the facility, or the manner in which

the plant is operated and surveilled, the

proposed changes do not create the

possibility of a new or different kind of

accident from any previously analyzed.

3. The proposed changes do not

involve a significant reduction in a

margin of safety.

There is no adverse impact on

equipment design or operation and

there are no changes being made to the

Technical Specification required safety

limits or safety system settings that

would adversely affect plant safety. The

proposed changes are administrative in

nature and do not change the level of

programmatic controls and procedural

details associated with the

aforementioned surveillance

requirements.

From the evaluations performed on

the subject SSCs there are no

indications that potential problems

would be cycle-length dependent or that

potential degradation would be

significant for the time frame of interest

and, therefore, increasing the

surveillance interval to the bounding

limit of 30 months (24 months plus

25%) will have little, if any, impact on

safety.

The proposed changes to the

surveillance intervals are still consistent

with the basis for the intervals and the

intent and method of performing the

surveillance is unchanged. Deletion of

the restriction ‘‘during shutdown’’

where this restriction is stated will

permit performance of certain

maintenance and testing activities

during conditions or modes other than

shutdown. North Atlantic will ensure,

through the implementation of

appropriate administrative controls, that

proper regard to their effect on safe

operation of the plant is given prior to

conduct of a particular surveillance in a

condition or mode other than shutdown.

In addition, use of the subject SSCs

during normal plant operation,

combined with their previous history of

availability and reliability, provide

assurance that the proposed changes

will not affect the reliability of the

subject SSCs. Thus, it is concluded that

the subject SSCs would be available

upon demand to mitigate the

consequences of an accident and,

therefore, there is no 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: Exeter Public Library,

Founders Park, Exeter, NH 03833.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, CT 06141–0270

NRC Project Director: Cecil O.

Thomas.

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50–336,

Millstone Nuclear Power Station, Unit

No. 2, New London County, Connecticut

Date of amendment request: April 29,

1998.

Description of amendment request:

The proposed amendment would allow

NNECO to revise the Updated Final

Safety Analysis Report (UFSAR) for

Millstone Unit 2 by deleting the

diversity requirement for the two lowrange pressurizer pressure transmitters,

PT–103 and PT–103–1.

NNECO proposes to replace PT–103

and PT–103–1 with transmitters that are

more accurate in a post-accident

environment to provide assurance that

entry into shutdown cooling in a postaccident environment is not

compromised and to provide relief for

the reactor coolant system pressure/

temperature curves. NNECO further

indicates that only a single model series

of Rosemount transmitters meet the

revised design requirements and has

specifically requested to delete the

diversity requirement in the UFSAR,

Section 4.3.8.2.3, ‘‘Pressurizer

Pressure.’’ NNECO has determined that

this deviation from the current design

basis constitutes an unreviewed safety

question as defined in 10 CFR 50.59.

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 replacement of the low-range

pressurizer pressure transmitters with

non-diverse transmitters will reduce the

instrument uncertainties post—SBLOCA

[small break loss-of-coolant accident] or

MSLB [main steamline break]. The

probability of a post-accident

intersystem LOCA [loss-of-coolant

accident] as a result of aligning the SDC

[shutdown cooling] system to RCS

[reactor coolant system] pressures

beyond its design pressure is reduced

due to the reduced uncertainties of lowrange pressurizer pressure signals to the

SDC suction valve interlocks. The

reduced uncertainty associated with the

low-range pressurizer pressure

transmitters in a harsh environment will

not significantly reduce the probability

of previously evaluated accidents

relative to the use of the transmitter

signal as an input variable to the ICC

[inadequate core cooling] system, or to

the other functions of LTOP [low

temperature overpressure protection],

SIT [safety injection tank] interlock, and

Hot Shutdown Panel indication since

these are functions not required in postaccident design bases. With respect to

ICC, other parameters are available to

the operator to determine adequate

cooling of the core is taking place and
Federal Register / Vol. 63, No. 106 /Wednesday, June 3, 1998 / Notices 30267

saturation conditions are being

approached or are occurring.

The loss of diversity in manufacturer

and operating principle results in a

small increase in the susceptibility of

the replacement transmitters to common

cause events that are primarily linked to

internal failures of the transmitters

verses failure that result from external

events or conditions. To some extent,

externally related common cause

failures that can result from calibration

or maintenance errors can be expected

to also increase slightly because of

commonality of procedures. Common

cause failure increase is considered for

the identified functions and all

accidents. Because of the slight increase

in the probability of common cause

failures, the probability of exceeding

RCS pressure/temperature (P/T) curves

at temperatures [less than] 275°F is

slightly increased (assuming a common

cause failure of the replacement

transmitters that would result in

indicating a pressure lower than actual

RCS pressure). Also, the potential for

exceeding ASME Section III, Appendix

G, pressure/temperature limit curves on

cooldown and heatup is also slightly

increased due to the slight increase in

potential for common cause failure. This

small increase in the potential for

common cause failure will not

significantly affect the probability of

previously evaluated accidents. The

reasons for this are:

—Exceedance of P/T limit curves does

not in and of itself result in an

accident initiator.

—Internally caused common cause

failures are not expected to have a

significant impact on the overall

common cause potential of the

transmitters. Typically, the majority

of common cause potential is due to

external reasons. Further, many times

simultaneous internal failures of

instrumentation can be recognized by

direct comparison at which time

alternative means can be sought, if

available. Unless failure of the

replacement transmitters was

simultaneous and resulted in

consistent output signals, transmitter

failure would likely be recognized

before requiring LTOP.

—The narrow-range pressurizer

pressure loops (1500–2500 psia

[pounds per square inch absolute]) are

fully qualified Class 1E with

transmitters manufactured by a

different vendor. They provide a

check against the low-range

pressurizer pressure transmitters in

the overlapped range of 1500–1600

psia.

—The non-class 1E wide-range

pressurizer pressure loop (0–3000

psig [pounds per square inch gauge]),

although not environmentally

qualified, utilizes a qualified type

transmitter manufactured by a

different vendor and has been

demonstrated to be reliable. It

provides a check against the lowrange pressurizer pressure

transmitters in the overlapped range

of 0–1600 psia.

The probability of an intersystem

LOCA will not increase, due to the

multiple means of determining that RCS

pressure is beyond the SDC system

design pressure and the multiple

failures that would have to occur. All

other functions evaluated (ICC, Hot

Shutdown Panel and the SIT interlock)

would not increase the probability of a

previously evaluated accident.

Because the function and output of

the replacement transmitters is the same

as the existing transmitters and the

transmitter failure types has not

changed, the radiological consequences

of previously evaluated accidents are

not affected by the proposed change.

The exception to this occurs when

considering consequences of accidents

that result in a harsh environment

inside Containment and requiring SDC

for long term cooling. In these cases,

(SBLOCAs and MSLBs inside

containment) given that transmitter

accuracy is improved, the ability of

getting onto SDC improves. This allows

getting onto SDC more consistently. By

doing so, a reduction in the radiological

consequences of these accidents may be

improved.

Therefore, there is no significant

increase in the probability or

consequences of an accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The function and output of the

replacement transmitters are the same as

the existing transmitters, and the

transmitter failure types have not

changed.

Therefore, the change 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.

Improvement of the transmitters’

ability to provide an accurate

interpretation of RCS pressures in the

operating range of 0–1600 psia (postaccident harsh environment in

Containment) results in a positive

benefit in the ability to control

cooldown rates, establish SDC, and

operate at the proper RCS pressures.

However, the Margin of Safety is not

impacted when the original

transmitters/uncertainty calculations are

compared to the proposed replacement

transmitters/uncertainty calculations

with regard to RCS P/T curves.

Therefore, the change will not involve

a reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: 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. 50–286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of amendment request: June 25,

1997.

Description of amendment request:

The proposed amendment would

change the Indian Point 3 Technical

Specifications to allow the use of

zirconium alloy or stainless steel filler

rods in fuel assemblies to replace failed

or damaged fuel rods.

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:

Consistent with the criteria of 10 CFR

50.92, the enclosed application is

judged to involve no significant hazards

based on the following information:

(1) Does the proposed license

amendment involve a significant

increase in the probability or

consequences of an accident previously

analyzed?

Response:

The proposed changes modify the

technical specification only to the

extent that the reconstitution is

recognized as acceptable under limited

circumstances. Reconstitution is limited

to substitution of zirconium alloy or

stainless steel filler rods, and must be in

accordance with approved applications

of fuel rod configurations. Although
30268 Federal Register / Vol. 63, No. 106 /Wednesday, June 3, 1998 / Notices

these changes permit reconstitution to

occur without the need for a specific

technical specification change, use of an

approved methodology is required prior

to its application. Since the changes will

allow substitution of filler rods for

leaking, potentially leaking rods or

damaged rods, the changes may actually

reduce the radiological consequences of

an accident. It is noted that the specific

changes requested in this letter have

previously been found acceptable by the

NRC in GL [Generic Letter] 90–02,

Supplement 1. For these reasons, we

conclude that the changes will 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 will not create

the possibility of a new or different kind

of accident from any accident

previously evaluated because they will

only affect the assembly configuration

and can only be implemented if

demonstrated to meet current plant

requirements in accordance with an

NRC-approved methodology. The other

aspects of plant design, operation

limitations, and responses to events will

remain unchanged. It is noted that the

changes have previously been

determined acceptable by the NRC in

GL 90–02, Supplement 1.

(3) Does the proposed amendment

involve a significant reduction in a

margin of safety?

Response:

The proposed change will not involve

a reduction in a margin of safety

because the changes can only be

implemented if demonstrated to meet

current plant requirements in

accordance with an NRC-approved

methodology. It is noted that the

changes have previously been

determined acceptable by the NRC in

GL 90–02, Supplement 1.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10601.

Attorney for licensee: Mr. David

Blabey, 10 Columbus Circle, New York,

New York 10019.

NRC Project Director: S. Singh Bajwa.

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: April 24,

1998.

Description of amendment request:

The proposed amendment would

change Technical Specification (TS)

Section 3/4.3.1.1, ‘‘Reactor Protection

System Instrumentation,’’ TS Section

3/4.3.2.1, ‘‘Safety Features Actuation

System Instrumentation,’’ TS Section 3/

4.3.2.2, ‘‘Steam and Feedwater Rupture

Control System Instrumentation,’’ and

the associated TS bases. The TS tables

of response time limits would be

relocated to the Davis-Besse Technical

Requirements Manual. Other changes in

these TS sections consistent with the

relocation are also proposed.

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 (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 no

accident initiator, conditions or

assumptions are affected by the

proposed changes to Technical

Specification (TS) 3/4.3.1.1, Reactor

Protection System (RPS)

Instrumentation, TS 3/4.3.2.1, Safety

Features Actuation System (SFAS)

Instrumentation, and TS 3/4.3.2.2,

Steam and Feedwater Rupture Control

System (SFRCS) Instrumentation and

the associated TS Bases to relocate their

tables of response time limits to the

Technical Requirements Manual (TRM)

of the DBNPS Updated Safety Analysis

Report (USAR).

The RPS, SFAS and SFRCS response

time limits and surveillance intervals

currently prescribed in the TS are not

changed under the proposed License

Amendment. The RPS, SFAS and

SFRCS will continue to function in the

manner described in the DBNPS USAR.

Therefore, the performance of these

protection systems will remain within

the bounds of the USAR accident

analysis.

Under the proposed changes, the

response time limits of the RPS, SFAS

and SFRCS would continue to be tested

in accordance with plant procedures in

the same manner as in the past. The

specific RPS, SFAS and SFRCS tables of

response time limits will be relocated

and remain controlled by the TRM of

the DBNPS USAR following the

guidance of the NRC’s Generic Letter

(GL) 93–08, ‘‘Relocation of Technical

Specification Tables of Instrument

Response Time Limits,’’ dated

December 29, 1993. Any change to the

relocated tables for response time limits

will be subject to review and evaluation

under Section 50.59, ‘‘Changes, Tests,

and Experiments,’’ of Title 10 of the

Code of Federal Regulation (10 CFR)

prior to any changes being made.

1b. Not involve a significant increase

in the consequences of an accident

previously evaluated because no

accident conditions or assumptions are

affected by the proposed changes. As

described above, the changes are

consistent with the guidance of NRC GL

93–08. The proposed changes

administratively relocate response time

tables and do not alter the source term,

containment isolation, or allowable

releases. The proposed changes,

therefore, will not increase the

radiological consequences of a

previously evaluated accident.

2. Not create the possibility of a new

or different kind of accident from any

accident previously evaluated because

no new accident initiators or

assumptions are introduced by the

proposed changes, which involve only

the administrative relocation of

response time limit tables. No new

accident scenarios, transient precursors,

failure mechanisms, or limiting failures

are introduced as a result of the

proposed changes. As described above,

the changes are consistent with the

guidance of NRC GL 93–08. The

proposed changes do not alter any

accident scenarios and future changes to

the response time limits will be subject

to the regulatory requirements in 10

CFR 50.59.

3. Not involve a significant reduction

in a margin of safety because the

proposed changes only administratively

relocate the response time tables from

the TS to the USAR TRM, and do not

reduce or adversely affect the

capabilities of any plant structures,

systems or components. No response

times will be changed by this

amendment request. Future changes to

the response time limits will be subject

to the regulatory requirements of 10 CFR

50.59. Accordingly, there is not a

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three
Federal Register / Vol. 63, No. 106 /Wednesday, June 3, 1998 / Notices 30269

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, OH 43606.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Acting Project Director: Ronald

R. Bellamy.

Wisconsin Public Service Corporation,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request: May 14,

1998.

Description of amendment request:

The proposed Technical Specification

(TS) amendment would redefine the

parent tube pressure boundary location

for Westinghouse mechanical hybrid

expansion joint (HEJ) steam generator

(SG) tube sleeves. The proposed

amendment would change the parent

tube pressure boundary definition from

a minimum required interference lip to

a minimum required length of nondegraded hardroll engagement.

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 was reviewed in

accordance with the provisions of 10

CFR 50.92 to show no significant

hazards exist. The proposed change will

not:

(1) Involve a significant increase in

the probability or consequence of an

accident previously evaluated.

Mechanical testing shows inherent

structural integrity of the HEJ upper

joint such that the requirements of RG

1.121 are met even for 360 degree, 100

percent throughwall parent tube

indications (PTIs). Structural test results

are documented in WCAP–15050. Based

on the test data, the structural

recommendations of RG 1.121 are

satisfied when there is a minimum

length of non-degraded hardroll which

measures 0.92 inch (plus an allowance

for NDE measurement uncertainty) or

more from the bottom of the hardroll

upper transition (HRUT), as measured

on the inside of the sleeve. Based on the

structural integrity of the HEJ upper

joint, it can be concluded that

application of the revised parent tube

pressure boundary will not result in a

significant increase in the probability of

an accident previously evaluated.

A conservatively bounding primaryto-secondary steam line break (SLB) leak

rate of one gpm will be applied to the

calculation for postulated SLB leakage.

This leak rate encompasses all HEJs left

inservice with PTIs located outside the

revised parent tube pressure boundary.

This one gpm is based on a normal

operating leakage limit of 150 gpd. This

leak rate is based on tests and analysis

documented in WCAP–15050.

Application of this leak rate to the

postulated leakage calculation will

ensure primary-to-secondary leakage

will not exceed the current maximum

allowable during a SLB event.

Maintenance of the current maximum

allowable primary-to-secondary leak

rate during a SLB event ensures off-site

doses will not exceed a small fraction of

10 CFR 100 and control room doses will

not exceed GDC–19 criteria. Therefore,

it can be concluded that the application

of the revised parent tube pressure

boundary will not increase the

consequences of an accident previously

evaluated.

(2) Create the possibility of a new or

different kind of accident from any

previously evaluated.

Implementation of the revised parent

tube pressure boundary will not

introduce a change to the design basis

or operation of the plant. The

configuration of the currently installed

sleeves is not physically changed. As

with the initial installation of the

sleeves and previous changes to the

parent tube pressure boundary for HEJs,

implementation of the revised parent

tube pressure boundary does not

interact with other portions of the

reactor coolant system. Neither the

sleeve design nor the implementation of

the revised parent tube pressure

boundary affects any other component

or location of the tube outside of the

immediate repaired area. Mechanical

testing of representative specimens

supports the conclusions that the joint

retains structural integrity consistent

with RG 1.121 and leakage integrity

with regards to 10 CFR 100 and GDC–

19. Any hypothetical accident as a

result of potential PTIs is bounded by

the existing steam generator tube

rupture analysis. Therefore, application

of the revised parent tube pressure

boundary will not create the possibility

of a new or different kind of accident

from any previously evaluated.

(3) Involve a significant reduction in

the margin of safety.

The safety factors used in

establishment of the HEJ sleeved tube

pressure boundary are consistent with

safety factors in the ASME Boiler and

Pressure Vessel Code used in the SG

design. Based on the sleeve-to-tube

geometry, it is unrealistic to consider

that application of the revised parent

tube pressure boundary could result in

single tube leak rates exceeding the

normal makeup capacity during normal

operating conditions. The parent tube

pressure boundary developed in

WCAP–15050 has been developed using

the methodology of RG 1.121. The

performance characteristics of

postulated degraded parent of HEJ

sleeve/tube joints have been verified

through testing to retain structural

integrity and preclude significant

leakage during both normal operating

and SLB conditions. The existing offsite and control room dose evaluation

performed for KNPP established a

faulted loop primary-to-secondary leak

rate of 12.85 gpm. Combined leakage

from all sources including the assumed

leak rate for the voltage based repair

criteria and for HEJs with PTIs that are

left inservice will not exceed 12.85 gpm

in the faulted loop. Maintenance of this

limit will ensure off-site doses will not

exceed a small fraction of the 10 CFR

100 guidelines nor will it exceed the

GDC–19 criteria for control room dose.

Therefore, the application of the revised

parent tube pressure boundary will 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: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, WI 54311–7001.

Attorney for licensee: Bradley D.

Jackson, Esq., Foley and Lardner, P.O.

Box 1497, Madison, WI 53701–1497.

Acting NRC Project Director: Ronald

R. Bellamy.

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
30270 Federal Register / Vol. 63, No. 106 /Wednesday, June 3, 1998 / Notices

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

amendment, (2) the amendment, and (3)

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the

local public document rooms for the

particular facilities involved.

Carolina Power & Light Company, et al.,

Docket No. 50–325, Brunswick Steam

Electric Plant, Unit 1, Brunswick

County, North Carolina

Date of application for amendments:

February 23, 1998, as supplemented

March 27, 1998.

Brief description of amendment: The

amendment modifies the values for the

safety limit for the Minimum Critical

Power Ratio (SLMCPR) in the TS and

the associated action statement for Cycle

12 operation only. A reference in TS

6.9.3.2.c is also revised.

Date of issuance: May 11, 1998.

Effective date: May 11, 1998.

Amendment No.: 194.

Facility Operating License No. DPR–

71: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 10, 1998 (63 FR 17900).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 11, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297.

Commonwealth Edison Company,

Docket Nos. 50–373 and 50–374, LaSalle

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of application for amendments:

September 26, 1997, as supplemented

on April 7, 1998, and May 1, 1998.

Brief description of amendments: The

amendments revise the Technical

Specifications to upgrade the ventilation

filter testing program to the current

industry standards and specify that the

auxiliary electric equipment room is

required to be habitable during design

bases accidents. Revisions related to

drywell and suppression chamber purge

and the editorial changes requested in

the September 26, 1997, application

were approved and issued under

Amendment Nos. 125 and 110 dated

April 27, 1998.

Date of issuance: May 13, 1998.

Effective date: Immediately, to be

implemented prior to restart from L1F35

for Unit 1 and prior to restart from the

current outage for Unit 2.

Amendment Nos.: 126 and 111.

Facility Operating License Nos. NPF–

11 and NPF–18: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: November 19, 1997 (62 FR

61840). The April 7 and May 1, 1998,

submittals provided additional

information that did not change the

initial proposed no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated May 13, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348.

Consolidated Edison Company of New

York, Docket No. 50–247, Indian Point

Nuclear Generating Unit No. 2,

Westchester County, New York

Date of application for amendment:

July 8, 1994, as supplemented August

13, 1996, and February 12, 1998.

Brief description of amendment: The

amendment revises Technical

Specifications Sections 3.7 and 3.3.E to

clarify offsite power availability

requirement, revise emergency diesel

generator fuel oil availability

requirements and specify the

configuration requirements for removing

Component Cooling Pump 22 from

service.

Date of issuance: May 8, 1998.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 196.

Facility Operating License No. DPR–

26: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 17, 1994 (59 FR

42336).

The August 13, 1996, and February

12, 1998, letters provided clarifying

information that did not change the

initial proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 8, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

Consumers Energy Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan

Date of application for amendment:

January 18, 1996, as supplemented

October 1, 1997, and January 29 and

April 27, 1998.

Brief description of amendment: The

amendment revises the technical

specifications regarding inspection

requirements for the reactor coolant

pump (RCP) flywheels. The staff denied

a portion of the amendment request

regarding application to the flywheel

testing program of the Surveillance

Requirement 4.0.2 allowance for

surveillance interval extension of up to

25%. A separate Notice of Partial Denial

of Amendment to Facility Operating

License and Opportunity for Hearing

has been published in the Federal

Register.

Date of issuance: May 15, 1998.

Effective date: May 15, 1998.

Amendment No.: 182.

Facility Operating License No. DPR–

20: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 5, 1997 (62 FR

59915).

The January 29 and April 27, 1998,

letters provided additional clarifying

information that was within the scope of

the original Federal Register notice and

did not change the staff’s initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 15, 1998.

No significant hazards consideration

comments received: No.
Federal Register / Vol. 63, No. 106 /Wednesday, June 3, 1998 / Notices 30271

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Public Service Electric & Gas Company,

Docket No. 50–354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

December 19, 1997, as supplemented

March 6, 1998.

Brief description of amendment: This

amendment changes the wording of

Section 4.2.2, ‘‘Terrestrial Ecology

Monitoring,’’ of the Environmental

Protection Plan to include completion of

the Salt Drift Monitoring Program.

Date of issuance: May 8, 1998.

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 111.

Facility Operating License No. NPF–

57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 28, 1998 (63 FR 4321)

The March 6, 1998, supplement

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination or expand the scope of

the original Federal Register notice.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 8, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

Southern Nuclear Power Company, Inc.,

Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton,

Georgia, Docket Nos. 50–424 and 50–

425, Vogtle Electric Generating Plant,

Units 1 and 2, Burke County, Georgia

Date of application for amendments:

January 22, 1998, as supplemented by

letter dated March 18, 1998, April 21,

1998, and May 15, 1998.

Brief description of amendments: The

amendments change the Technical

Specifications to allow an extended

allowed outage time for one emergency

diesel generator of 14 days.

Date of issuance: May 20, 1998.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: Unit 1–100; Unit

2–78.

Facility Operating License Nos. NPF–

68 and NPF–81: Amendments revised

the Technical Specifications and

Operating Licenses.

Date of initial notice in Federal

Register: February 11, 1998 (63 FR

6998) The March 18, 1998, April 21,

1998, and May 15, 1998, supplements

provided clarifying information that did

not change the scope of the January 22,

1998, application and the initial

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated May 20, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia.

Tennessee Valley Authority, Docket

Nos. 50–260 and 50–296, Browns Ferry

Nuclear Plant , Units No. 2 and 3,

Limestone County, Alabama

Date of amendment request:

December 11, 1996, as supplemented by

letter dated November 3, 1997 (TS–386).

Description of amendment request:

The amendment modifies the Appendix

A Technical Specifications (TSs)

Limiting Safety System Setting 2.2.A,

which relates to the main steam safety/

relief valve set points and the set point

tolerance. Specifically, the revision

increases the set point tolerance to ±3%

vice the current ± 11 pound per square

inch (approximately 1% of set point

value) tolerance. Bases 1.2 and 3.6D/

4.6D also are revised.

Date of issuance: May 18, 1998.

Effective date: May 18, 1998.

Amendment No.: 251 and 210.

Facility Operating License Nos. DPR–

33, DPR–52, and DPR–68. Amendment

revised the TSs.

Date of initial notice in Federal

Register: January 15, 1997 (62 FR 2194).

The licensee’s letter dated November 3,

1997, provided additional supporting

information that did not change the

initial proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated May 18, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Athens Public Library, 405

South Street, Athens, Alabama 35611.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont

Date of application for amendment:

March 20, 1998

The licensee proposed to modify the

licensing basis by limiting the time the

large (18′′) purge and vent valves may be

open to 90 hours per year. This is a

change to the Final Safety Analysis

Report (FSAR) and technical

specification bases.

Date of Issuance: May 14, 1998.

Effective date: May 14, 1998.

Amendment No.: 161.

Facility Operating License No. DPR–

28: Amendment authorizes revision to

the FSAR.

Date of initial notice in Federal

Register: March 27, 1998 (63 FR 14976).

The Commission’s related evaluation

of this amendment is contained in a

Safety Evaluation dated May 14, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Dated at Rockville, Maryland, this 27th day

of May 1998.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects—

III/IV, Office of Nuclear Reactor Regulation.

[FR Doc. 98–14519 Filed 6–2–98; 8:45 am]

BILLING CODE 7590–01–P

OFFICE OF PERSONNEL

MANAGEMENT

Submission for OMB Review;

Comment Request for Revision of

Information Collection, RI 20–64 & RI

20–64A

AGENCY: Office of Personnel

Management.

ACTION: Notice.

SUMMARY: In accordance with the

Paperwork Reduction Act of 1995 (Pub.

L. 104–13, May 22, 1995), this notice

announces that the Office of Personnel

Management will submit to the Office of

Management and Budget a request for

revision of the following information

collection. RI 20–64, Former Spouse

Survivor Annuity Election, is used by

the Civil Service Retirement System to

provide information about the amount

of annuity payable after a survivor

reduction and obtain a survivor benefits

election form from annuitants who are

eligible to elect to provide survivor

benefits for a former spouse. RI 20–64A,

Information On Electing A Survivor

Annuity For Your Former Spouse, is a

pamphlet that provides important

information to retirees under the Civil

Service Retirement System who want to

provide a survivor annuity for a former

spouse.

Approximately 30 RI 20–64 forms are

completed annually. Each form takes

about 45 minutes to complete. The

annual estimated burden is 23 hours.

For copies of this proposal, contact

Jim Farron on (202) 418–3208, or E-mail

to jmfarron@mail.opm.gov