61836 Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices

only; no specific action or written

response is required. Conformance with

the guidance provided in the generic

letter is voluntary.

The generic letter is available in the

NRC Public Document Room under

accession number 9711050091.

DATES: The generic letter was issued on

November 13, 1997.

ADDRESSEES: Not applicable.

FOR FURTHER INFORMATION CONTACT: Beth

A. Wetzel at (301) 415–1355.

SUPPLEMENTARY INFORMATION:

Addressees of GL 96–06 have

experienced difficulty in determining

and implementing corrective actions for

resolving the issues identified in the

generic letter. Additionally, questions

have been raised regarding (1) the risk

implications of installing relief valves to

deal with the thermal

overpressurization issue; (2) the use of

the ASME Code, Section III, Appendix

F, criteria for permanent resolution of

the thermal overpressurization issue;

and (3) the NRC staff’s closure of

Generic Safety Issue 150,

‘‘Overpressurization of Containment

Penetrations.’’ Given these

considerations, risk insights, and

industry initiatives that are being

considered or that may be proposed,

addressees may require additional time

to fully evaluate and resolve the GL 96–

06 issues. Therefore, addressees who

find it necessary to revise their

corrective actions or schedular

commitments for resolving GL 96–06

issues may submit a revised response to

the generic letter. Nevertheless, specific

corrective actions that have been

defined and are clearly needed should

not be delayed without suitable

justification.

Dated at Rockville, Maryland, this 13th day

of November 1997.

For the Nuclear Regulatory Commission.

David B. Matthews,

Acting Director, Division of Reactor Program

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 97–30332 Filed 11–18–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Sunshine Act Meeting

DATE: Weeks of November 17, 24,

December 1 and 8, 1997.

PLACE: Commissioners’ Conference

Room, 11555 Rockville Pike, Rockville,

Maryland.

STATUS: Public and Closed.

MATTERS TO BE CONSIDERED:

Week of November 17

Friday, November 21

11:30 a.m. Affirmation Session (Public

Meeting)

A. Final Rule—Deliberate Misconduct

by Unlicensed Persons (Tentative)

B. Louisiana Energy Services—

Financial Qualifications Aspects of

Petitions for Review of LBP–96–25

(Contact: Ken Hart, 301–415–1659)

Week of November 24—Tentative

There are no meetings the week of

November 24.

Week of December 1—Tentative

There are no meetings the week of

December 1.

Week of December 8—Tentative

Thursday, December 11

2:00 p.m. Briefing on Investigative

Matters (Closed—Ex. 5 & 7)

3:00 p.m. Affirmation Session (Public

meeting) (if needed)

Friday, December 12

9:00 a.m. Meeting with Northeast

Nuclear on Millstone (Public

meeting) (Contact: Bill Travers,

301–415–1200)

Note: 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://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, DC 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: November 14, 1997.

William M. Hill, Jr.,

SECY Tracking Officer, Office of the

Secretary.

[FR Doc. 97–30527 Filed 11–17–97; 12:32 pm]

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 October 27,

1997, through November 6, 1997. The

last biweekly notice was published on

November 5, 1997 (62 FR 59912).

Notice Of Consideration Of Issuance Of

Amendments ToFacility 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
Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices 61837

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 Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555-0001, and

should cite the publication date and

page number of this Federal Register

notice. Written comments may also be

delivered to Room 6D22, Two White

Flint North, 11545 Rockville Pike,

Rockville, Maryland from 7:30 a.m. to

4:15 p.m. Federal workdays. Copies of

written comments received may be

examined at the NRC Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing

of requests for a hearing and petitions

for leave to intervene is discussed

below.

By December 19, 1997, the licensee

may file a request for a hearing with

respect to issuance of the amendment to

the subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

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.
61838 Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices

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 amendments request: October

22, 1997

Description of amendments request:

The proposed amendment incorporates

both steady state and transient degraded

voltage setpoints into Technical

Specifications, as opposed to the current

single degraded voltage setpoint. The

proposed changes ensure adequate

terminal voltage to all safety-related

equipment during steady state and

transient voltage conditions.

Additionally, the 4 kV voltage range

required during testing of the emergency

diesel generators (EDGs) will be

decreased to ensure the new steady state

degraded voltage relays are not actuated

during testing and to ensure the 4 kV

motors are operated within their voltage

rating.

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. Would not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The proposed changes revise the current

degraded voltage setpoint and adds an

additional steady state undervoltage

requirement to Unit 1 and 2 Technical

Specifications. The current degraded voltage

relays will be referred to as

≥transient degraded voltage relays.’’ The

new settings allow for calibration tolerances,

potential transformer correction factors, test

equipment uncertainties, and relay drift. The

nominal settings account for the above

factors, plus additional margin to the

analytical limit. The acceptable voltage range

during EDG surveillance testing is also being

decreased. The setpoint and time delay

associated with the 4 kV bus loss of voltage

relays is unaffected by this amendment

request.

The accident analyses credit the loading of

the EDGs based on loss of offsite power. The

4 kV emergency bus loss of voltage and

degraded voltage relays initiate starting and

loading of the emergency diesel generators

(EDGs) when the preferred power source

voltage is lost or drops below a

predetermined value. The relays also initiate

disconnection of the preferred power source

from the 4 kV emergency busses. These

actions ensure adequate terminal voltage to

all safety-related electrical equipment

required to support accident mitigation. The

required voltage necessary to ensure safetyrelated motors are capable of starting is 75

percent of nominal rated equipment voltage.

The required voltage necessary to ensure

these motors continue running for extended

periods is 90 percent of nominal rated

equipment voltage.

The degraded (transient) voltage setpoint is

being changed from 3628 [plus or minus] 25

Volts to 3710 [plus or minus] 80 Volts. Based

on the most recent calculations, a minimum

voltage of 3630 Volts is required to ensure at

least 75 percent of the nominal voltage is

available to No. 13 Charging Pump, which is

the most limiting electrical load.

The new steady state degraded voltage

relay setpoint will be established at 3900

[plus or minus] 80 Volts. The setpoint

ensures that there is at least 90 percent of

nominal voltage available to No. 13 Charging

Pump. The time delay associated with this

actuation is 101 [plus or minus] 3.5 seconds.

The time delay provides adequate time for

the voltage regulator to recover bus voltage

following a voltage swing on the 500 kV

system and time for the EDG voltage

regulator to stabilize. The steady state

degraded voltage relays will be tested in the

same manner, and at the same frequency, as

the loss of voltage and transient degraded

voltage relays.

The required voltage range during EDG

surveillance testing is being revised from

4160 [plus or minus] 420 Volts to 4160 +240,

-100 Volts. The surveillance requirement

verifies that the EDG voltage regulator is

maintaining an acceptable voltage. The

new value ensures the 4 kV motors are

operated within their rated voltage and

prevents actuation of the steady state

degraded voltage relay during surveillance

testing.

The degraded voltage relays are not

initiators in any previously evaluated

accidents. Additionally, decreasing the

acceptable voltage range during EDG testing

does not affect the initiation of any

previously analyzed accidents. Therefore, the

proposed changes do not involve an increase

in the probability or consequences of an

accident previously analyzed.

2. Would not create the possibility of a new

or different type of accident from any

accident previously evaluated.

The license amendment request revises the

current degraded voltage setpoint and adds

an additional steady state degraded voltage

requirement. Additionally, the acceptable

voltage range during EDG surveillance testing

is being decreased. The proposed changes

ensure adequate starting and running

terminal voltage to all safety-related electrical

equipment during steady state and transient

degraded voltage conditions. The addition of

the steady state degraded voltage relays

provide an extra scheme of protection against

sustained degraded voltage conditions. The

facility currently relies upon degraded

voltage relays to start and load the EDGs and

to disconnect the preferred power source

from the 4 kV emergency busses. Therefore,

revising the relay setpoint, adding additional

steady state degraded voltage protection, and

decreasing the acceptable voltage range

during EDG testing does not create the

possibility of a new or different type of

accident from any accident previously

evaluated.

3. Would not involve a significant

reduction in a margin of safety.

The safety function of the degraded voltage

relays is to ensure that the preferred power

source is disconnected from the 4 kV

emergency busses during loss of voltage or

degraded voltage conditions. The relays also

ensure the EDGs are started and loaded.

Ultimately, these actions ensure the minimal

terminal voltage necessary to start and run all

safety-related electrical equipment is

maintained. The proposed changes revise the

current degraded voltage setpoint and adds

an additional steady state undervoltage

requirement. Additionally, the acceptable

voltage range during EDG surveillance testing

is being decreased to ensure actuation of the

steady state degraded voltage relays does not

occur during

EDG testing, and to ensure the 4 kV motors

are operated within their rated voltage range.

Because the proposed changes ultimately

ensure adequate terminal voltage to all

safety-related electrical equipment during

transient and steady state undervoltage

conditions, the safety function of the

degraded voltage relays, as well as the margin

of safety afforded by these relays is

unchanged.

Therefore, the 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 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendments request

involves no significant hazards

consideration.

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678

Attorney for licensee: Jay E. Silbert,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: S. Singh Bajwa,

Director

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: October

2, 1997

Description of amendment request:

The proposed amendment would

address an unreviewed safety question

associated with the analysis of a fuel

handling accident in the Fuel Storage

Building as described in Section 15.7.4,

‘‘Design Basis Fuel Handling

Accidents,’’ of the H.B. Robinson Steam

Electric Plant (HBR) Updated Final

Safety Analysis Report (UFSAR).

Carolina Power & Light Company (the

licensee) determined that an assumption

used in the accident analysis for depth

of water above the top of irradiated fuel

in the spent fuel pit was nonconservative. The accident analysis

assumed a depth of 23 feet instead of

the correct value of 21 feet. The licensee

has submitted a revised accident

analysis using the correct assumption

and has proposed that the UFSAR be
Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices 61839

changed to incorporate the results of the

revised analysis.

Basis for proposed no significant

hazards 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 the UFSAR is to

change assumptions associated with the

evaluation of a fuel handling accident in the

Fuel Storage Building. The change in

assumptions is to reduce the

decontamination factor associated with the

removal of elemental iodine from the spent

fuel pool water. Because the decontamination

factor for elemental iodine is reduced, the

consequences of a fuel handling accident in

the Fuel Storage Building is [sic] increased.

However, because the radiological

consequences remain well within the

exposure guideline values of 10 CFR 100,

paragraph 11 (i.e., 25% or less of the values),

the increase in consequences is not

significant. The change in assumptions for

the fuel handling accident in the Fuel Storage

Building do [sic] not affect operation,

maintenance, or design of equipment

associated with the handling of fuel in the

Fuel Storage Building, therefore, the

probability of a fuel handling accident as

previously evaluated is not changed.

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, changes in

parameters governing normal plant

operation, or methods of operation. The

proposed change does not introduce a new

mode of operation or changes in the method

of normal plant operation. Therefore, the

possibility of a new or different kind of

accident from any accident previously

evaluated is not created.

3. Does this change involve a significant

reduction in a margin of safety?

The proposed change to the UFSAR to

change the assumptions associated with a

fuel handling accident in the Fuel Storage

Building is to change the assumption for the

decontamination factor for elemental iodine

to a smaller value. The new assumption for

elemental iodine decontamination factor

preserves the approximate factor of 24

margin between experimental data for

elemental iodine decontamination factor and

the assumed value provided in NRC Safety

Guide 25. Therefore, the 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: R. E. Jones,

General Counsel, Carolina Power &

Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: James E.

LyonsCommonwealth Edison Company,

Docket Nos. 50-237 and 50-249, Dresden

Nuclear Power Station, Units 2 and 3,

Grundy County, Illinois Docket Nos. 50-

254 and 50-265, Quad Cities Nuclear

Power Station, Units 1 and 2, Rock

Island County, Illinois Date of

application for amendment request:

September 30, 1997

Description of amendment request:

This request changes the Technical

Specifications (TS) by adding a new

Section 3/4.12.C, ‘‘Inservice Leak and

Hydrostatic Testing Operation,’’ to

allow certain reactor coolant pressure

tests to be performed in MODE 4 when

the metallurgical characteristics of the

reactor pressure vessel require the

pressure testing at or approaching

temperatures ≤212°F, which normally

correspond with MODE 3.

Basis for proposed no significant

hazards 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 amendment represents the

addition of a Special Test Exception to

perform Pressure Testing Operations

consistent with the requirements of Section

3.10.1 of the Improved Standard Technical

Specifications (NUREG-1433). The proposed

changes are consistent with the current plant

safety analyses. Implementation of these

changes will provide continued assurance

that specified parameters associated with

Pressure Testing Operations will remain

within their acceptance limits, and as such,

will not significantly increase the probability

or consequences of a previously evaluated

accident.

The proposed changes are based on

requirements specified by Section 3.10.1 of

NUREG-1433. Any such changes are

consistent with the current plant safety

analyses and have been determined to

represent sufficient requirements for the

assurance and reliability of equipment

assumed to operate in the safety analyses, or

provide continued assurance that specified

parameters associated with Pressure Testing

Operations remain within their acceptance

limits. As such, these changes will not

significantly increase the probability or

consequences of a previously evaluated

accident.

The associated systems affecting Pressure

Testing Operations related to this proposed

amendment are not assumed in any analyses

to initiate any accident sequence; therefore,

the probability of any accident previously

evaluated is not increased by this proposed

amendment which incorporates the

requirements of Section 3.10.1 of NUREG1433. In addition, the proposed limiting

conditions for operation and surveillance

requirements for the proposed amendment

ensure a level of equipment operability

sufficient to mitigate any operational

occurrences which could occur while

operating under this Special Test Exception.

Furthermore, any operational occurrence

postulated during operation under this

Special Test Exception is bounded by the

Design Basis Accidents. Therefore, the

proposed amendment does not increase the

consequences of nay accident previously

evaluated.

There is no change to the consequences of

an accident previously evaluated because

Pressure Testing Operations does not

adversely affect either the on-site or off-site

does consequences resulting from an

accident. In addition, Pressure Testing

Operations is not an accident initiator. As

such, there is no adverse impact on the

probability of accident initiators. Thus, there

is no significant increase in the probability of

any previously analyzed accident.

2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated because:

The proposed amendment represents the

conversion of current Technical Specification

requirements to maintain consistency with

those requirements specified in Section

3.10.1 of NUREG-1433. The proposed

changes are consistent with the current plant

safety analyses. These proposed changes do

not involve revisions to the design of the

station. In addition, the proposed limiting

conditions for operation and surveillance

requirements for the proposed amendment

ensure a level of equipment operability

sufficient to mitigate any operational

occurrences which could occur while

operating under the Special Test Exception.

Some of the changes may involve revision in

the testing of components at the station;

however, these are in accordance with the

current plant safety analyses. The proposed

changes will not introduce new failure

mechanisms beyond those already

considered In the current plant safety

analyses.

The associated systems that affect Pressure

Testing Operations related to the proposed

amendment, are not assumed in any plant

safety analysis to initiate any accident

sequence. In addition, the proposed

surveillance requirements for any such

affected systems are consistent with the

requirements of Section 3.10.1 of NUREG1433. Therefore, the possibility of a new or

different kind of accident from any accident

previously evaluated is not created.

3) Involve a significant reduction in the

margin of safety because:

ComEd proposes to revise the Technical

Specifications to be consistent with those

provisions specified in Section 3.10.1 of

NUREG-1433. The proposed changes are

consistent with the current plant safety

analyses. In addition, these proposed changes

do not involve revisions to the design of the

station. As such, the proposed individual

changes will maintain the same level of
61840 Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices

reliability of the equipment associated with

Pressure Testing Operations, assumed to

operate in the plant safety analysis, or

provide continued assurance that specified

parameters affecting, will remain within their

acceptance limits. Therefore, the proposed

changes provide continued assurance of

Pressure Testing Operations without

adversely affecting the public health and

safety and as such, will not significantly

reduce existing plant safety margins.

The proposed amendment to the Technical

Specifications implements present

requirements, or the requirements in

accordance with the guidelines set forth in

Section 3.10.1 of NUREG-1433. The proposed

changes have been evaluated and found to be

acceptable for use at the stations based on

system design, safety analysis requirements,

and operational performance. Since the

proposed changes are based on NRC accepted

provisions that are applicable at the stations

and maintain necessary levels of system or

component reliability affecting Pressure

Testing Operations, 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 are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: for Dresden, Morris Area

Public Library District, 604 Liberty

Street, Morris, Illinois 60450; for Quad

Cities, Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021

Attorney for licensee: Michael I.

Miller, Esquire; Sidle and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: Robert A. Capra

Commonwealth Edison Company,

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

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of amendment request:

September 26, 1997

Description of amendment request:

The proposed amendments would

revise the Technical Specifications to

(1) prohibit the simultaneous opening of

the drywell and suppression chamber

purge system isolation valves, (2)

upgrade the ventilation filter testing

program to the latest industry standards,

and (3) specify that the auxiliary electric

equipment room is required to be

habitable during design bases accidents.

Basis for proposed no significant

hazards 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:

a. Drywell and Suppression Chamber Purge

System

The purpose of the drywell and

suppression chamber purge system isolation

valves is to mitigate the consequences of a

design bases accident. Operation of these

valves will have no effect on the probability

of a design bases accident occurring.

The current TS 3.6.1.8 allows for the

drywell and suppression chamber purge

system isolation valves to be open

simultaneously. In this condition,

containment pressure and offsite dose during

design bases accidents would be greater than

previously evaluated. The proposed revision

to TS 3.6.1.8 would prevent the simultaneous

opening of the drywell and suppression

chamber purge system isolation valves thus

assuring that the consequences of design

bases accidents previously evaluated are still

bounding.

b. Ventilation Filter Testing Program

The SBGTS [Standby Gas Treatment

System] and Control Room and AEER

[Auxiliary Electric Equipment Room]

Emergency Filtration Systems are designed to

mitigate the radiological consequences of

previously evaluated design bases accidents.

Operation and testing of these systems will

have no effect on the probability of a design

bases accident occurring.

The proposed revisions associated with

this change relocate the requirements for

SBGTS and Control Room and AEER

Emergency Filtration System filter testing

from the current TS SRs to a new TS

administrative control program. The testing

requirements are being upgraded to the latest

industry standards. Filter testing in

accordance with the proposed program will

ensure that Title 10, Code of Federal

Regulations, Part 50 (10 CFR 50), Appendix

A, General Design Criteria (GDC) 19 and 10

CFR 100 limits are not exceeded.

c. Other Control Room and Auxiliary

Electric Equipment Room Emergency

Filtration System Changes

The SBGTS and Control Room and AEER

Emergency Filtration System are designed to

mitigate the radiological consequences of

previously evaluated design bases accidents.

Operation and testing of these systems will

have no effect on the probability of a design

bases accident occurring.

The proposed revisions associated with

this change acknowledge that the AEERs are

required to be habitable during design bases

accidents. This is consistent with the

plant—s design bases.

d. Editorial Changes

The proposed revisions to TS 6.2.F.7

reformat the requirement to establish

consistency with the remainder of TS 6.2.F.

There are no technical changes being

proposed.

Based upon the above, the proposed

amendment will not increase the probability

or consequences of any accident previously

evaluated.

2) Create the possibility of a new or

different kind of accident from any accident

previously evaluated because:

a. Drywell and Suppression Chamber Purge

System

No new plant equipment is being installed,

and use of currently installed plant

equipment is not affected by this proposed

change. The proposed revision to TS 3.6.1.8

provides additional limitations on the

opening of the drywell and suppression

chamber purge system isolation valves.

b. Ventilation Filter Testing Program

No new plant equipment is being installed,

and use of currently installed plant

equipment is not affected by this proposed

change. These proposed revisions will

demonstrate operability of the Control Room

and AEER Emergency Filtration System using

the latest industry standards.

c. Other Control Room and Auxiliary

Electric Equipment Room Emergency

Filtration System Changes

No new plant equipment is being installed,

and use of currently installed plant

equipment is not affected by this proposed

change. These proposed revisions will

demonstrate habitability of the AEER by

imposing operability requirements on the

AEER recirculation filter units.

d. Editorial Changes

The proposed revisions to TS 6.2.F.7

reformat the requirement to establish

consistency with the remainder of TS 6.2.F.

There are no technical changes being

proposed.

Based upon the above, the proposed

change will not create the possibility of a

new or different kind of accident or transient

previously evaluated.

3) Involve a significant reduction in the

margin of safety because:

a. Drywell and Suppression Chamber Purge

System

The current TS 3.6.1.8 requirements are

non-conservative with respect to the

assumptions used when evaluating steam

bypass of the suppression chamber;

specifically, a maximum allowable leakage

area of 0.03 square feet with the only credible

leakage path was assumed to be suppression

chamber vacuum breaker valve seat leakage.

This proposed revision to TS 3.6.1.8 will

make the TS requirements consistent with

those assumptions.

b. Ventilation Filter Testing Program

These proposed revisions will ensure

operability of the Control Room and

Auxiliary Electric Equipment Room (AEER)

Emergency Filtration system using the latest

industry standards. Filter testing in

accordance with the proposed program will

ensure that GDC 19 and 10 CFR 100 limits

are not exceeded.

c. Other Control Room and Auxiliary

Electric Equipment Room Emergency

Filtration System Changes

These proposed revisions will ensure

operably of the control room and AEER

Emergency Filtration System by

demonstrating system performance with the

control room and AEER recirculation filter

units to ensure GDC 19 limits are not

exceeded.

d. Editorial Changes

The proposed revisions to TS 6.2.F.7

reformat the requirement to establish

consistency with the remainder of TS 6.2.F.

There are no technical changes being

proposed.
Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices 61841

Based on the above, the proposed TS

change does not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location:Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: Robert A. Capra

Commonwealth Edison Company,

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

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of amendment request: October

15, 1997

Description of amendment request:

The proposed amendments would

revise Technical Specification Table

4.3.7.5-1, Accident Monitoring

Instrumentation Surveillance

Requirements, by deleting a footnote

that provides details concerning the

calibration requirements for the drywell

hydrogen concentration analyzer and

monitor.

Basis for proposed no significant

hazards 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:

The drywell hydrogen concentration

analyzer and monitors are required to be

operable by TS 3/4.7.5, Accident Monitoring

Instrumentation. Table 4.3.7.5-1, Accident

Monitoring Instrumentation Surveillance

Requirements, includes a footnote providing

unnecessary details related to the calibration

of this specific analyzer and monitors. The

footnote provides information that was

determined to put the hydrogen analyzers

and monitors outside of the design basis by

limiting the range of the indication to 0% to

4% hydrogen in the drywell. The calibration

method is being corrected to provide the

correct range of 0% to 10%, and requires this

note in the TS to be changed or deleted. The

footnote is proposed to be deleted from the

TS, because it provides unnecessary detail.

Deletion of the footnote will not cause an

increase in the probability of an accident,

because this instrumentation is only for

accident monitoring instrumentation and

thus does not affect accident initiators or

assumptions.

Deletion of the footnote will not change the

consequences of an accident previously

evaluated, because this detail in the TS does

not change the requirement of performing a

channel calibration at the specified

frequency. In addition, the ability to monitor

hydrogen during an accident will not be

affected by deletion of the footnote.

Therefore, this change 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 because:

This is monitoring instrumentation only.

Deletion of the footnote concerning specifics

on how to calibrate this instrumentation will

not affect the reliability or failure modes of

the drywell hydrogen concentration analyzer

and monitors. Therefore, this 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 the

margin of safety because:

This is monitoring instrumentation only.

Deletion of the footnote concerning specifics

on how to calibrate this instrumentation will

not change the requirement to perform

Channel Calibrations at the frequency

specified in the TS. The details of how to

perform a Channel Calibration on the drywell

hydrogen concentration analyzer and

monitors are located in plant procedures and

are in accordance with vendor

recommendations. The TS requirements for

redundancy of the instrumentation and the

actions to be taken for inoperable

instrumentation are also not affected by the

deletion of this footnote.

This change to the level of information

regarding this calibration is consistent with

the detail for this and other instrumentation

in NUREG-1434, Revision 1, Standard

Technical Specifications, General Electric

Plants, BWR/6.

Therefore, deletion of footnote * from TS

Table 4.3.7.5-1 will not involve a reduction

in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location:Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: Robert A. Capra

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: October

6, 1997

Description of amendment request:

The proposed amendments would

delete all references to the steam line

low pressure safety injection function.

Basis for proposed no significant

hazards 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?

Answer

Probability

Accident initiators can affect the

probability of a previously evaluated

accident. The addition of a new device or

piece of equipment to the plant may

introduce a new accident initiator. No new

equipment is added to the plant as a result

of this change. The proposed removal of the

low steam line steam pressure will involve

removing the steam line pressure safety

injection function. This results in a reduction

in the likelihood of spurious safety

injections. Spurious safety injections can

result in inadvertent ECCS [emergency core

cooling system] actuations. Inadvertent ECCS

Actuation is a UFSAR [updated Final Safety

Analysis Report] accident (UFSAR 15.5.1).

Therefore, this change will result in a

reduction in the probability of an accident

previously evaluated.

Routine plant operating practices and

conditions will not be altered by the removal

of the safety injection function. Therefore,

there is no operating practice or condition

change that could increase the probability of

occurrence of a previously evaluated

accident.

There is no significant increase in the

probability of an accident previously

evaluated.

Consequences

Accidents previously evaluated that could

be adversely affected are the steam line break

and the feedwater line break. These accidents

will result in secondary side depressurization

with pressure reaching the current actuation

setpoint. The review of these accidents found

that the consequences of the previous

accident analysis acceptance criteria remain

satisfied. The specifics of the accident

analysis is discussed below.

The steam line break accident was

analyzed to demonstrate short term cooling

capability. A spectrum of break sizes were

evaluated to determine the limiting break

size. For smaller breaks (including the

limiting break size), the safety injection

actuation on low pressurizer pressure occurs

prior to low steam line pressure safety

injection. However, for larger steam line

breaks the setpoint for low steam line

pressure safety injection is reached prior to

low pressurizer pressure safety injection. The

larger spectrum of breaks were analyzed

without credit for the low steam line pressure

safety injection. The results of this analysis

found that there would be a slight increase

in time required for safety injection to

actuate. The low pressurizer safety injection

would actuate in these accidents due to the

cooldown and depressurization of the reactor

coolant system in response to the secondary

side energy removal. The Departure from
61842 Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices

Nucleate Boiling Ratios (DNBRs) were

analyzed with this time delay in safety

injection. The DNBRs for these cases were

found to be less limiting than those

calculated for the limiting break size.

Therefore, the removal of steam line low

pressure safety injection does not adversely

affect the DNBR, fuel failure or dose

consequences of the main steam line break

accident. Other acceptance criteria would not

be expected to be affected by the small

change in timing of the safety injection

signal.

In addition, to the Chapter 15 accident

analysis, the Chapter 6 containment response

to mass and energy releases was evaluated

without credit for steam line low pressure

safety injection. The evaluation demonstrated

that for steam line breaks inside of

containment, the high containment pressure

safety injection set point is reached prior to

the pressure associated with steam line low

pressure safety injection. Therefore the

existing containment response evaluation is

not adversely affected by the removal of the

low steam pressure safety injection. This also

assures that the existing environmental

qualification envelope for McGuire is not

affected by this change. For steam line breaks

outside of containment the maximum

required breaksize is 1.0 ft2, which results in

transients with safety injection caused by low

pressurizer pressure prior to low steam line

pressure safety injection.

The feedwater line break accidents were

analyzed to demonstrate long term core

cooling capability. During a feedwater line

break, the secondary system will

depressurize if the break occurs between the

main feedwater check valve and the steam

generator. However, breaks are only required

to occur at the terminal ends of feedwater

piping (i.e., at the feedwater pump or at the

steam generator). For a feedwater line break

at the main feedwater pump, the main feed

check valve will prevent depressurization of

the steam generator. For a feedwater line

break at the steam generator, a safety

injection on high containment pressure will

occur prior to safety injection on steam

pressure. Therefore, the elimination of the

steam line low pressure safety injection does

not adversely impact the feedwater line break

accident.

In summary, a review was conducted of all

design basis accidents to identify those

which result in a low steam pressure safety

injection. These accidents were then

evaluated to verify that the accident analysis

were within acceptance criteria. This review

revealed that all accident analysis results

were within current analysis acceptance

criteria.

Therefore, there is no significant increase

in the consequences of a previously

evaluated accident.

Conclusion

Elimination of the low steam line pressure

safety injection results in no 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[?]

Answer

There is no introduction of new equipment

or operating practices that could result in a

new operating condition. The plant will

continue to operate in the same method with

the same complement of equipment with the

exception of the actuation logic associated

with the steam line low pressure safety

injection. Therefore, there is no new

operating condition that would be expected

to generate a new sequence of events which

could generate a new or different accident.

There is no new equipment that could

interact with other plant structures, systems

or components.

The low pressure safety injection

equipment is the only plant equipment

affected by this change. There are no new

equipment failure modes which might result

in a new or different accident.

Affected accidents were evaluated to

validate that the accident sequence would

not deviate in a fashion which would create

a new or different accident. The analysis of

the feedwater line break and steam line break

did not reveal any new or different type of

accident.

Removal of the low steam line pressure

safety injection will not 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?

Answer

The margin of safety relevant to this

change is represented by the margin of

physical protection provided by fuel

cladding and the reactor containment. Effects

of this change on the safety analysis was

described under question 1 above. The

results of the analysis demonstrate that

DNBR, fuel clad integrity and containment

response were not significantly affected by

the removal of low steam line pressure safety

injection. Therefore, the physical protection

provide[d] by the fuel cladding and reactor

containment were not affected by this

change. Accident acceptance criteria

continued to be met without credit for the

safety function. The radiological

consequences of accidents was not affected

by the change.

The removal of the low steam line pressure

safety injection did not significantly reduce

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: J. Murrey Atkins Library,

University of North Carolina at

Charlotte, 9201 University City

Boulevard, North Carolina

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50-423,

Millstone Nuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of amendment request: October

15, 1997

Description of amendment request:

The proposed amendment would affect

nominal trip setpoints and allowable

values for Reactor Trip System (RTS)

Instrumentation Trip Setpoints Table

2.2-1, and Engineered Safety Features

Actuation System (ESFAS)

Instrumentation Trip Setpoints Table

3.3-4. In addition, the proposed

amendment would (1) decrease the

reactor trip setpoint for the reactor

coolant pump (RCP) low shaft speed

(underspeed trip setpoint) from 95.8

percent to 92.4 percent of rated speed,

(2) make editorial changes, and (3)

change the Bases to reflect the new

methodology.

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration, which

is presented below:

NNECO has reviewed the proposed

revision in accordance with 10CFR50.92 and

has concluded that the revision does not

involve a significant hazards consideration

(SHC). The basis for this conclusion is that

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

satisfied. The proposed revision does not

involve [an] SHC because the revision would

not:

1. Involve a significant increase in the

probability or consequence of an accident

previously evaluated.

The proposed changes to Tables 2.2-1 and

3.3-4 involve changes from a five column

format to a two column format. The RTS trip

setpoints and ESFAS trip setpoints remain

unchanged with the exception of the RCP

low shaft speed trip setpoint discussed

below. Detailed operability criteria will be

moved to surveillance procedures and

analysis has demonstrated that an adequate

margin for normal trip setpoints exist and

safety analysis limits are preserved in all

RTS/ESFAS functions.

Changing the RCP low shaft speed trip

setpoint will not change the probability of

occurrence of the event. The existing

accident analysis (Millstone Unit No. 3 FSAR

[final safety analysis report] section 15.3.2) of

the complete loss of forced reactor coolant

flow remains valid for the proposed change.

Therefore, the change to the RCP low shaft

speed trip setpoint does not increase the

probability or consequences of any

previously analyzed accident.

In addition, the proposed changes to

Tables 2.2-1 and 3.3-4 do not alter the intent

or method by which the surveillances are

conducted. Therefore, the scope of evaluation

performed gives reasonable assurance that

there will not be an adverse impact on the

consequences or the probability of any

previously analyzed accident.

Therefore, the proposed revision does not

involve a significant increase in the

probability or consequence of an accident

previously evaluated.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.
Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices 61843

The existing design basis adequately covers

the plant response with the proposed change

to the RCP low shaft speed trip setpoint. The

change does not introduce new failure

modes.

The proposed changes to Tables 2.2-1 and

3.3-4 do not modify the design or operation

of any plant system. The proposed changes

do not alter the intent or method by which

the surveillances are conducted, other than

adjusting the allowable values to reflect

historical instrument performance data.

Therefore, the proposed revision does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The proposed changes to Tables 2.2-1 and

3.3-4 modify the existing five column format

to a two column format to show the RTS and

ESFAS nominal trip setpoints and the

process rack bistable allowable values for

individual functions. Detailed operability

criteria will be moved to the surveillance

procedures. With the exception of the low

shaft speed trip discussed below, the RTS

and ESFAS setpoints remain unchanged and

analysis has demonstrated that an adequate

margin for normal trip setpoints exist and

safety analysis limits are preserved in all

RTS/ESFAS functions.

Since the safety limits of the design are

still met, the proposed change to the RCP low

shaft speed trip setpoint does not reduce the

margin of safety.

Therefore, the proposed revision does not

involve a significant reduction in a margin of

safety.

In conclusion, based on the information

provided, it is determined that the proposed

revision does not involve an SHC.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut

NRC Deputy Director: Phillip F. McKee

Pacific Gas and Electric Company,

Docket Nos. 50-275 and 50-323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of amendment requests: March

10, 1997, as supplemented by letter

dated May 20, 1997

Description of amendment requests:

The proposed amendments would

revise the combined Technical

Specifications (TS) for the Diablo

Canyon Power Plant Unit Nos. 1 and 2

to revise TS 3/4.4.5 and 3.4.6.2,

including associated Bases 3/4.4.5 and

3/4.4.6.2, to allow the implementation

of steam generator (SG) tube alternate

repair criteria for axial indications in

the Westinghouse explosive tube

expansion (WEXTEX) region below the

top of the tubesheet and below the

bottom of the WEXTEX transition that

may exceed the current TS depth-based

plugging limit. The allowed primary-tosecondary operational leakage from any

one SG would be reduced from 500 gpd

to 150 gpd.

Basis for proposed no significant

hazards 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.

Probability

Of the various accidents previously

evaluated, the proposed changes only affect

the steam generator tube rupture (SGTR)

event evaluation and the postulated steam

line break (SLB) accident evaluation. Loss-ofcoolant accident (LOCA) conditions cause a

compressive axial load to act on the tube.

Therefore, since the LOCA tends to force the

tube into the tubesheet rather than pull it out,

it is not a factor in this amendment request.

Another faulted load consideration is a safe

shutdown earthquake (SSE); however, the

seismic analysis of Series 51 steam generators

has shown that axial loading of the tubes is

negligible during an SSE.

For the SGTR event, the required structural

margins of the steam generator tubes will be

maintained by the presence of the tubesheet.

Tube rupture is precluded for cracks in the

Westinghouse explosive tube expansion

(WEXTEX) region due to the constraint

provided by the tubesheet. Therefore,

Regulatory Guide (RG) 1.121, ‘‘Bases for

Plugging Degraded PWR Steam Generator

Tubes,’’ margins against burst are maintained

for both normal and postulated accident

conditions.

The W* length supplies the necessary

resistive force to preclude pullout loads

under both normal operating and accident

conditions. The contact pressure results from

the WEXTEX expansion process, thermal

expansion mismatch between the tube and

tubesheet and from the differential pressure

between the primary and secondary side.

Therefore, the proposed change results in no

significant increase in the probability of the

occurrence of an SGTR or SLB accident. 1

The proposed changes do not affect other

systems, structures, components or

operational features. Therefore, based on the

above evaluation, the proposed changes do

not involve a significant increase in the

probability of an accident previously

evaluated.

Consequences

The consequences of an SGTR event are

affected by the primary-to-secondary leakage

flow during the event. Primary-to-secondary

leakage flow through a postulated broken

tube is not affected by the proposed change

since the tubesheet enhances the tube

integrity in the region of the WEXTEX

expansion by precluding tube deformation

beyond its initial expanded outside diameter.

The resistance to both tube rupture and

collapse is strengthened by the tubesheet in

that region. At normal operating pressures,

leakage from primary water stress corrosion

cracking (PWSCC) in the W* length is limited

by both the tube-to-tubesheet crevice and the

limited crack opening permitted by the

tubesheet constraint. Consequently,

negligible normal operating leakage is

expected from cracks within the tubesheet

region.

SLB leakage is limited by leakage flow

restrictions resulting from the crack and tubeto-tubesheet contact pressures that provide a

restricted leakage path above the indications

and also limit the degree of crack face

opening compared to free span indications.

The total leakage, that is, the combined

leakage for all such tubes, plus the combined

leakage developed by any other ARC, must be

below the maximum allowable SLB leak rate

limit, such that off-site doses are maintained

less than 10 CFR 100 guideline values.

Therefore, based on the above evaluation,

the proposed changes do not involve a

significant 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 do not introduce

any changes or mechanisms that create the

possibility of a new or different kind of

accident. Tube bundle integrity is expected

to be maintained for all plant conditions

upon implementation of the proposed steam

generator alternate tube plugging criteria.

WCAP-14797, Revision 1, ‘‘Generic W*

Tube Plugging Criteria for 51 Series Steam

Generator Tubesheet Region WEXTEX

Expansions,’’ requires that any tubes with

indications identified using the bobbin coil

probe during the bobbin sampling plan also

be inspected with the RPC coil throughout

the W* length of the tubes. The use of the

RPC will: (a) identify any new or nonexpected degradation mode that may not be

identified using the bobbin coil probe, and

(b) confirm and characterize the bobbin coil

indication.

These changes do not introduce any new

equipment or any change to existing

equipment. No new effects on existing

equipment are created nor are any new

malfunctions introduced.

Therefore, based on the above evaluation,

the proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes maintain the

required structural margins of the steam
61844 Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices

generator tubes for both normal and accident

conditions. RG 1.121 is used as the basis in

the development of the W* alternate tube

plugging criteria for determining that steam

generator tube integrity considerations are

maintained within acceptable limits. RG

1.121 describes a method acceptable to the

NRC staff for meeting General Design Criteria

14, 15, 31, and 32 by reducing the probability

and consequences of an SGTR. RG 1.121

concludes that by determining the limiting

safe conditions of tube wall degradation

beyond which tubes with unacceptable

cracking, as established by inservice

inspection, should be removed from service

or repaired, the probability and consequences

of an SGTR are reduced. This RG uses safety

factors on loads for tube burst that are

consistent with the requirements of Section

III of the ASME Code.

For primarily axially oriented cracking

located within the tubesheet, tube burst is

precluded due to the presence of the

tubesheet. WCAP-14797 defines a length,

W*, of degradation free expanded tubing that

provides the necessary resistance to tube

pullout due to the pressure induced forces

(with applicable safety factors applied).

Application of the W* criteria will preclude

unacceptable primary-to-secondary leakage

during all plant conditions. The methodology

for determining leakage provides for large

margins between calculated and actual

leakage values in the W* criteria.

Plugging of the steam generator tubes

reduces the reactor coolant flow margin for

core cooling. Implementation of the proposed

changes are expected to result in plugging of

fewer tubes than with the current criteria.

Thus, implementation of the proposed

changes will maintain the margin of flow that

may have otherwise been reduced by tube

plugging.

Based on the above, it is concluded that the

proposed changes do not result in a

significant reduction of margin with respect

to plant safety as defined in the FSAR Update

or bases of the plant Technical

Specifications.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407

Attorney for licensee: Christopher J.

Warner, Esq., Pacific Gas and Electric

Company, P.O. Box 7442, San

Francisco, California 94120

NRC Project Director: William H.

Bateman

Philadelphia Electric Company, Docket

No. 50-352, Limerick Generating

Station, Unit 1, Montgomery County,

Pennsylvania

Date of amendment request: October

24, 1997

Description of amendment request:

The proposed Technical Specifications

(TS) changes would revise TS Section 3/

4.1.3.6 to exempt control rod 50-27 from

the coupling test for the remainder of

Cycle 7.

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration, which

is presented below:

1. The proposed Technical Specifications

(TS) change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated. The

probability of occurrence of the analyzed

Control Rod Drop Accident (CRDA) is not

increased by operating with the subject

control blade in a condition not known to be

coupled since the compensatory measures

will assure that the blade will remain fully

inserted below 10% rated thermal power

where the CRDA is a concern. Monitoring of

nuclear instrumentation responses in the

vicinity of the blade when the drive is

withdrawn above 10% power will assure the

blade is tracking with the drive with no

potential to stick and then drop. Scram

impact forces from an uncoupled control rod

are of insufficient energy to dislodge the fuel

support (or fuel) or to cause a threat to the

pressure boundary integrity. No reduction of

system or equipment redundancy is

involved.

The CRDA analyzed in the Safety Analysis

Report (SAR) remains the limiting rod drop

accident, and its consequences are unaffected

by operation of the subject blade in the

proposed manner. Operation of the control

blade as described, i.e., withdrawn no further

than the 46 position and in a condition not

known to be coupled, has no adverse effect

on scram performance in response to any

other postulated accident. The scram insert

motion of the rod is not affected by the

potentially uncoupled condition, and since

the rod is already partially inserted at

position 46, it should have a slightly better

negative reactivity insertion characteristic.

Therefore, no potential to increase onsite or

offsite radiological consequences beyond

those previously analyzed in the SAR is

created.

Operating the subject control blade in a

condition not known to be coupled does not

result in any onsite or offsite radiological

consequences different from those previously

analyzed in the SAR. The subject control

blade will be fully inserted below 10%

thermal power where the CRDA is a concern

and will be monitored during drive

withdrawal above 10% thermal power to

assure it is tracking with the drive. Scram

performance is not adversely affected by

operation from the near full-out position of

46. Hence, no new failure modes are created

and consequences of any postulated failures

are not increased.

Therefore, the proposed TS change does

not involve an increase in the probability or

consequences of an accident previously

evaluated.

2. The proposed TS change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The Safety Analysis Report (SAR) analyzed

Control Rod Drop Accident (CRDA) remains

the only type of accident initiated (or

contributed to) by the control rod drive/

control blade interface. The compensatory

actions to be taken when operating the

subject blade in a condition not verified to

be coupled assure that no new types of

accidents can occur. The subject control

blade will be fully inserted below 10%

thermal power where the CRDA is a concern

and will be monitored during drive

withdrawal above 10% thermal power to

assure it is tracking with the drive. Scram

performance is not adversely affected by

operation from the near full-out position of

46. Since no adverse effect on insertion or

scram performance is expected, the

previously analyzed accidents encompass

any potential consequence of operating with

an uncoupled control blade.

The compensatory actions to be taken

when operating the subject blade in a

condition not verified to be coupled assure

that no new failure modes are created, and,

therefore, no new type of equipment

malfunction is introduced by operating the

subject control blade in the proposed

manner.

Therefore, the proposed TS change does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed TS change does not

involve a significant reduction in a margin of

safety.

Operation with the subject control blade in

a condition not known to be coupled for the

remainder of Cycle 7 at LGS [Limerick

Generating Station] Unit 1, but with the

compensatory actions described below, does

not reduce the existing margin of safety

determined by the analysis of the Control

Rod Drop Accident (CRDA). The CRDA

analyzed in the Safety Analysis Report (SAR)

remains bounding in that the subject rod will

be fully inserted below 10% rated thermal

power where the CRDA is a concern. Above

10% power, when the associated drive is

withdrawn, the nuclear instrumentation in

the vicinity of the blade will be monitored to

assure the blade tracks with the drive,

providing assurance that the position of the

blade can be ascertained by the drive

position. If the control blade can not be

verified to have followed the drive, then the

rod shall be completely inserted and the

control rod directional valves disarmed in

accordance with existing TS requirements.

To minimize any scram impact loadings, the

blade will be operated at the near full-out

position of 46 except for intermediate

positions temporarily occupied during

standard rod withdrawal sequences.

Operating the subject control blade in the

proposed manner will have no adverse effect

on insertion or scram performance of the

blade and will preserve the margin of safety.
Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices 61845

Therefore, the proposed TS change does

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:Pottstown Public Library, 500

High Street, Pottstown, PA 19464

Attorney for licensee: J. W. Durham,

Sr., Esquire, Sr. V. P. and General

Counsel, Philadelphia Electric

Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company,

Docket No. 50-272, Salem Nuclear

Generating Station, Unit No. 1, Salem

County, New Jersey

Date of amendment request: October

6, 1997

Description of amendment request:

The amendment to the Technical

Specifications would increase the

allowable band for control and

shutdown rod demanded position

versus indicated position from plus or

minus 12 steps to plus or minus 18

steps when the power level is not

greater than 85% rated thermal power.

The amendment is identical to

Amendment 183 for Salem Unit 2,

which was issued September 10, 1997,

as an exigent amendment.

Basis for proposed no significant

hazards 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 to the rod

misalignment criteria of [plus or minus] 18

steps for core powers equal to or below 85%

of RATED THERMAL POWER (RTP) does not

increase the probability of previously

evaluated accidents. Increasing the

magnitude of the allowed control rod

misalignment is not a contributor to the

mechanistic cause of an accident evaluated

in any accident analysis. The magnitude of

control rod indicated misalignment is a

parameter used to establish the initial

conditions for accident evaluation.

The proposed increase in the allowable rod

misalignment from the current [plus or

minus] 12 steps for reactor powers equal to

or less than 85% RTP does not involve a

significant increase in the consequence of

any previously evaluated accident. Rod

misalignment affects power distribution,

shutdown margin and the ejected rod

accident. An extension of the allowable rod

misalignment above and below 85% RTP has

been analyzed in Westinghouse WCAP14672. As provided in WCAP-14672, above

85% the allowable misalignment is governed

by the available peaking factor margins as

determined by flux maps.

[Public Service Electric & Gas] PSE&G is

simplifying the proposed change by keeping

the currently allowed [plus or minus] 12 step

misalignment in Technical Specifications

3.1.3.1 and 3.1.3.2.1 for reactor power greater

than 85% RTP.

The PSE&G proposed change is to allow

[plus or minus] 18 steps misalignments in

Technical Specifications 3.1.3.1 and 3.1.3.2.1

for reactor power less than or equal to 85%

RTP. As demonstrated in WCAP-14672, for

reactor powers less than 85% RTP, the

available peaking factor margin increases

faster than any penalty associated with a

[plus or minus] 18 step misalignment.

As described in Section 4.0 of the

Westinghouse WCAP, a conservative penalty

factor has been applied to the rod insertion

allowance (RIA) of the shutdown margin

calculation to account for rods misaligned an

additional [plus or minus] 6 steps (for a total

of [plus or minus] 18 steps). This

conservative penalty factor is applied as part

of the reload analysis in order to satisfy

Technical Specification 3.1.1.1.

In addition to the normal, or Condition 1,

operational transients, the impacts of

increased rod misalignment on Condition II,

III and IV accident analysis have also been

evaluated. The proposed increase in rod

misalignment does not have a significant

effect on any moderator or Doppler reactivity

coefficients or defects, boron worth or reactor

kinetics parameters.

To account for the potential increase in

ejected rod parameters, conservative penalty

factors have been applied to the reload safety

evaluation to cover the additional [plus or

minus] 6 step misalignment. Margin is

available in the reload safety analysis to

accommodate this impact.

Therefore, the proposed amendment does

not increase 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.

0No new accident scenarios, failure

mechanisms or limiting single failures are

introduced as a result of the proposed change

to the rod misalignment criteria of [plus or

minus] 18 steps below 85% RTP. The

implementation of the proposed rod

misalignment criteria will have no adverse

effect on the performance of any other safety

related system. Therefore, the proposed

amendment does not create the possibility of

a new or different kind of accident from any

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in the margin

of safety. The Technical Specifications

allowed increase in peaking factors as power

is reduced accommodates the peaking factor

penalty associated with the additional [plus

or minus] 6 step misalignment for core

powers equal to or less than 85% RTP.

Therefore, there is no change to the peaking

factors assumed in the safety analysis. In

addition to peaking factors, there is no

change in any other current limit input into

the safety analysis. As the input, or initial

conditions, of the safety analysis have not

changed, there is no reduction in the margin

to 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

Public Service Electric & Gas Company,

Docket No. 50-272, Salem Nuclear

Generating Station, Unit No. 1, Salem

County, New Jersey

Date of amendment request: October

14, 1997

Description of amendment request:

The proposed amendment will modify

the Salem Unit 1 Technical

Specification (TS) 3.4.6.3, ‘‘Primary

Coolant System Pressure Isolation

Valves Limiting Condition for

Operation,’’ to be consistent with Salem

Unit 2 TSs.

Basis for proposed no significant

hazards 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 majority of the proposed changes, as

described above, are editorial in nature.

Rewording, and reformatting the Limiting

Condition for Operation, including the

surveillance requirements do not involve a

significant increase to the probability or

consequences of an accident.

Those substantive changes involving the

addition of (1) new reactor coolant system

pressure isolation valves, (2) providing for a

shorter test frequency upon entry into Mode

4, and (3) adding a new surveillance test

requirement, do not increase the probability

or consequences of an accident. These

changes ensure that the system and

components needed to prevent and minimize

the effects of inter-system loss of coolant are

properly identified in the Technical

Specifications.

Although pressure isolation valves are

being added to the Technical Specification

table, these valves were already included in

the IST [inservice testing] program as

pressure isolation valves and were being

tested as such. The proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.
61846 Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices

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, as described above,

does not physically alter the facility or the

operation of the facility. The majority of the

changes are editorial in nature and provide

for improvement in the human factors of the

Technical Specifications, while properly

identifying all the pressure isolation valves

in the Technical Specifications. The addition

of valves into the Technical Specification is

an administrative change that improves the

quality of the LCO [limiting condition for

operation], but does not add components to

the facility.

The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The margin of safety, as defined in the

bases for any technical specifications,

depend upon proper identification of

equipment and performance of the proper

surveillance requirements to demonstrate

equipment operability. The proposed change

will ensure that the proper valves are

identified and tested in accordance with the

Technical Specification requirements.

The proposed changes do not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: 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

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: October

21, 1997

Description of amendment request:

The proposed amendment revises

Technical Specification Tables 3.3-1

and 4.3-1 to require that Functional

Unit, ≥2. Power Range, Neutron Flux,’’

be operable in Mode 3, as well as in

Modes 1 and 2. The change is being

proposed because the licensee has

determined that the power range

nuclear instrumentation should be

operable in Mode 3 whenever the

reactor trip system breakers are in the

closed position and the control rods are

capable of being withdrawn.

Basis for proposed no significant

hazards 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 requirement for operability of a trip

and the surveillance requirements to ensure

the functionality of the trip are independent

of the probability of an accident previously

evaluated. The accident that this trip is

intended to mitigate is the Rod Withdrawal

from Subcriticality event. The surveillance

procedure and the requirement for the trip to

be operational when the Control Rod Drive

System is capable of rod movement mitigate

the consequences of this event, and do not

increase the probability of a rod withdrawal

from subcritical.

Therefore, the probability and

consequences of an accident previously

evaluated are not significantly increased.

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 do not involve any

modifications to existing plant equipment, do

not alter the function of any plant systems,

do not introduce any new operating

configurations or new modes of plant

operation, or change the safety analyses. The

proposed change is intended to ensure that

the trip function is available and will

perform as designed in the event of a

previously evaluated event.

The proposed changes 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 proposed change does not reduce the

margin of safety, because assurance of the

operability of the trip function is increased

by the proposed change.

Based on the above, PSE&G [Public Service

Electric & Gas Company] has determined that

the proposed changes do 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: Salem Free Public library, 112

West Broadway, Salem, NJ

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

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: August

28, 1997

Description of amendment request:

The proposed change would revise

Technical Specification 3.4.11, ‘‘Reactor

Coolant System (RCS) Pressure and

Temperature (P/T) Limits,’’ to

incorporate the new P/T curves, which

were provided by General Electric

Nuclear Energy in report number GENE-B1301793-01, ‘‘Perry Unit 1 RPV

Surveillance Materials Testing and

Analysis.’’

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration which

is presented below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change will provide for

approved P/T limit curves which are valid

through 9 effective full-power years (EFPY)

and 18 EFPY. This change will not affect any

Safety Limits, Power Distribution Limits, or

Limiting Conditions for Operation. The

proposed changes incorporate operating

limits which provide margin to brittle failure

of the reactor vessel based on testing of the

irradiated reactor vessel materials (base

metal, weld material, and heat affected zone

material). The limits ensure that adequate

safety margins against nonductile or rapidly

propagating failure exist during normal

operation, anticipated operational

occurrences, and system hydrostatic tests.

The specimens have been tested and

analyzed in accordance with 10 CFR 50,

Appendices G and H, using the methods

described in Generic Letter 88-11 and

Regulatory Guide 1.99 Revision 2. The

predicted lowest upper shelf energy at 32

EFPY was greater than the minimum

required by 10 CFR 50, Appendix G. The

adjusted reference temperature for the

limiting material was lower than the 200

degree Fahrenheit limit required by

Regulatory Guide 1.99 Revision 2. As such,

the integrity of the reactor pressure coolant

boundary is maintained. The changes will

result in equivalent or more conservative

limits on reactor vessel pressure as a function

of temperature for all operational conditions

(hydrostatic and leak testing, non-nuclear

heatup/cooldown, and core critical

operations). The methodology used to derive

these values produces limits which continue

to ensure that sufficient margin is maintained

to meet the criteria of GDC 31, ‘‘Fracture

Prevention of Reactor Coolant Pressure

Boundary.’’ There are no plant modifications

associated with this change and no new or

revised system interfaces. The proposed
Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices 61847

changes do not increase the probability of

occurrence or consequences previously

evaluated because the temperature shifts are

well within equipment operating ranges. As

such, there is no increase in the probability

of occurrence or 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 do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated because the proposed changes do

not involve any new modes of operation. The

only change will be operation of the plant

within operating pressure limits which are

determined in a more conservative manner.

Therefore, no new failure mode or accident

sequence is introduced by this change.

The testing and analysis meets 10 CFR 50,

Appendices G and H, requirements;

therefore, no new accident types, such as

brittle fracture of a reactor pressure coolant

boundary component is postulated. The

adjusted reference temperature and upper

shelf energy predicted at 32 EFPY are well

within the limits of 10 CFR 50, Appendices

G and H. Therefore, the possibility of an

accident of a new or different type than any

previously evaluated is not created.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The P/T limits are established to provide

acceptable margins for the operation of the

reactor coolant system during heat up and

cool down, criticality, and hydrotest

conditions. Technical Specification 3.4.11

limits the rates of change of temperature and

pressure to values consistent with the

fracture toughness requirements of 10 CFR

50, Appendices G and H, and ASME Boiler

and Pressure Vessel Code Section III

Appendix G. The bases section for Technical

Specification 3.4.11 refers to 10 CFR 50,

Appendices G and H, and ASME Code

Section III Appendix G. Changes in these

limits are necessary because the fracture

toughness properties of ferritic materials in

the reactor vessel change as a function of

reactor operating time. The specific

requirements for fracture toughness and

reactor vessel material surveillance that must

be considered in developing the P/T limits

are defined by 10 CFR 50, Appendices G and

H. The specific limits defined by 10 CFR 50,

Appendices G and H, set the margin of safety

for the reactor pressure vessel coolant

boundary. Since the testing and analysis of

the vessel specimens meet the requirements

and limits defined in 10 CFR 50, Appendices

G and H, the margin of safety as defined in

the basis for Technical Specification 3.4.11 is

not reduced. The revised curves are based on

the latest NRC guidelines along with actual

neutron fluence data for Perry. The new

limits conservatively account for irradiation

embrittlement effects, thereby maintaining

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: Gail H. Marcus

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:

September 8, 1997

Description of amendment request:

The proposed amendment would revise

Technical Specification (TS) 5.2.2.e,

‘‘Organization - Unit Staff,’’ by removing

the reference to the NRC Policy

Statement on working hours.

Administrative procedures will be

developed to limit the working hours of

unit staff who perform safety-related

functions.

Basis for proposed no significant

hazards 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 to TS 5.2.2.e only

alters the administrative location of and the

regulatory controls applicable to unit staff

specific overtime limits and working hours.

Overtime will remain controlled by plant

administrative procedures. Changes to the

relocated overtime limits and working hours

will be subject to review and evaluation

under 10 CFR 50.59, ‘‘Changes, Tests and

Experiments.’’ There is not an 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.

There is not an increase in the radiological

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. Therefore, there is no

increase in the radiological consequences of

an accident previously evaluated.

2. The proposed change does not create the

possibility or a new or different kind of

accident from any accident previously

evaluated.

The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated because the proposed change does

not change 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 change. No new or different types

of failures or accident initiators are

introduced by the proposed change.

The proposed change to TS 5.2.2.e only

alters the administrative location of and the

regulatory controls applicable to unit staff

specific overtime limits and working hours.

Therefore, there is no possibility created for

a new or different kind of accident.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change does not involve a

reduction in a margin of safety because unit

staff overtime is not an input in the

calculation of a safety margin with regard to

Technical Specification Safety Limits,

Limiting Safety System Settings, other

Technical Specification Limiting Conditions

for Operation, the Operational Requirements

Manual, or other previously defined margins

for any structure, system, or component

important to safety. The proposed change to

TS 5.2.2.e only alters the administrative

location of and the regulatory controls

applicable to unit staff specific overtime

limits and working hours.

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: Gail H. Marcus

TU Electric Company, Docket Nos. 50-

445 and 50-446, Comanche Peak Steam

Electric Station, Units 1 and 2,

Somervell County, Texas

Date of amendment request: October

24, 1997

Brief description of amendments:

Change to the core safety limit curves

and overtemperature N-16 reactor trip

function setpoints to support operation

with Unit 1, cycle 7 core configuration.

Basis for proposed no significant

hazards determination: As required by

10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no

significant hazards consideration, which

is presented below:

1. Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

A. Revision to the Unit 1 Core Safety

Limits
61848 Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices

Analyses of reactor core safety limits are

required as part of reload calculations for

each cycle. TU Electric has performed the

analyses of the Unit 1, Cycle 7 core

configuration to determine the reactor core

safety limits. The methodologies and safety

analysis values result in new operating

curves which, in general, permit plant

operation over a similar range of acceptable

conditions. This change means that if a

transient were to occur with the plant

operating at the limits of the new curve, a

different temperature and power level might

be attained than if the plant were operating

within the bounds of the old curves.

However, since the new curves were

developed using NRC approved

methodologies which are wholly consistent

with and do not represent a change in the

Technical Specification BASES for safety

limits, all applicable postulated transients

will continue to be properly mitigated. As a

result, there will be no significant increase in

the consequences, as determined by accident

analyses, of any accident previously

evaluated.

B. Revision to Unit 1 Overtemperature N16 Reactor Trip Setpoints

As a result of changes discussed, the

Overtemperature reactor trip setpoint has

been recalculated. These trip setpoints help

ensure that the core safety limits are

protected and that all applicable limits of the

safety analysis are met.

Based on the calculations performed, no

significant changes to the safety analysis

values for Overtemperature reactor trip

setpoint were required. The f(deltaI) trip reset

function was revised due to more top-skewed

axial power distributions predicted for this

cycle. The analyses performed show that,

using the TU Electric methodologies, all

applicable limits of the safety analysis are

met. This setpoint provides a trip function

which allows the mitigation of postulated

accidents and has no impact on accident

initiation. Therefore, the changes in safety

analysis values do not involve an increase in

the probability of an accident and, based on

satisfying all applicable safety analysis

limits, there is no significant increase in the

consequences of any accident previously

evaluated.

In addition, sufficient operating margin has

been maintained in the overtemperature

setpoint such that the risk of turbine

runbacks or reactor trips due to upper

plenum flow anomalies or other operational

transients will be minimized, thereby,

reducing potential challenges to the plant

safety systems.

SUMMARY

The changes in the amendment request

applies NRC approved methodologies to

changes in safety analysis values, new core

safety limits and new N-16 setpoint and

parameter values to assure that all applicable

safety analysis limits have been met. The

potential for an operational transient to occur

has not been affected and there has been no

significant impact on the consequences of

any accident previously evaluated.

2. Do the proposed changes create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

The proposed changes involve the

calculation of new reactor core safety limits

and overtemperature reactor trip setpoint

resets. As such, the changes play an

important role in the analysis of postulated

accidents but none of the changes effect plant

hardware or the operation of plant systems in

a way that could initiate an accident.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Do the proposed changes involve a

significant reduction in a margin of safety?

In reviewing and approving the methods

used for safety analyses and calculations, the

NRC has approved the safety analysis limits

which establish the margin of safety to be

maintained. While the actual impact on

safety is discussed in response to question 1,

the impact on margin of safety is discussed

below:

A. Revision to the Unit 1 Reactor Core

Safety Limits

The NRC-approved TU Electric reload

analysis methods have been used to

determine new reactor core safety limits. All

applicable safety analysis limits have been

met. The methods used are wholly consistent

with Technical Specification BASES 2.1

which is the bases for the safety limits. In

particular, the curves assure that for Unit 1,

Cycle 7, the calculated DNBR is no less than

the safety analysis limit and the average

enthalpy at the vessel exit is less than the

enthalpy of saturated liquid. The acceptance

criteria remains valid and continues to be

satisfied; therefore, no change in a margin of

safety occurs.

B. Revision to Unit 1 Overtemperature N16 Reactor Trip Setpoints

Because the reactor core safety limits for

CPSES Unit 1, Cycle 7 are recalculated, the

Reactor Trip System instrumentation setpoint

values for the Overtemperature N-16 reactor

trip setpoint which protect the reactor core

safety limits must also be recalculated. The

Overtemperature N-16 reactor trip setpoint

helps prevent the core and Reactor Coolant

System from exceeding their safety limits

during normal operation and design basis

anticipated operational occurrences. The

most relevant design basis analysis in

Chapter 15 of the CPSES Final Safety

Analysis Report (FSAR) which is affected by

the Overtemperature reactor trip setpoint is

the Uncontrolled Rod Cluster Control

Assembly Bank Withdrawal at Power (FSAR

Section 15.4.2). This event has been analyzed

with the new safety analysis value for the

Overtemperature reactor trip setpoint to

demonstrate compliance with event specific

acceptance criteria. Because all event

acceptance criteria are satisfied, there is no

degradation in a margin of safety.

The nominal Reactor Trip System

instrumentation setpoints values for the

Overtemperature N-16 reactor trip setpoint

(Technical Specification Table 2.2-1) are

determined based on a statistical

combination of all of the uncertainties in the

channels to arrive at a total uncertainty. The

total uncertainty plus additional margin is

applied in a conservative direction to the

safety analysis trip setpoint value to arrive at

the nominal and allowable values presented

in Technical Specification Table 2.2-1.

Meeting the requirements of Technical

Specification Table 2.2-1 assures that the

Overtemperature reactor trip setpoint

assumed in the safety analyses remains valid.

The CPSES Unit 1, Cycle 7 Overtemperature

reactor trip setpoint is not significantly

different from the previous cycle, and thus

provides operational flexibility to withstand

mild transients without initiating automatic

protective actions. Although the value of the

f(deltaI) trip reset function setpoint is

different, the Reactor Trip System

instrumentation setpoint values for the

Overtemperature N-16 reactor trip setpoint

are consistent with the safety analysis

assumptions which have been analytically

demonstrated to be adequate to meet the

applicable event acceptance criteria. Thus,

there is no reduction in a margin of safety.

Using the NRC approved TU Electric

methods, the reactor core safety limits are

determined such that all applicable limits of

the safety analyses are met. Because the

applicable event acceptance criteria continue

to be met, there is no 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 Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019

Attorney for licensee: George L. Edgar,

Esq., Morgan, Lewis and Bockius, 1800

M Street, N.W., Washington, DC 20036

NRC Project Director: James W.

Clifford, Acting

Previously Published Notices Of

Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.
Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices 61849

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: October

13, 1997

Description of amendment request:

The proposed amendments would

support replacement of the three safetyrelated wide range level instruments.

The engineered safety features trip

setpoint for the refueling water

automatic switchover to recirculation

would be revised to account for the

difference in instrument uncertainty

associated with wide range level

instruments and provide additional

response time margin.

Date of publication of individual

notice in Federal Register: October 22,

1997 (62 FR 54859)

Expiration date of individual notice:

November 21, 1997

Local Public Document Room

location: J. Murrey Atkins Library,

University of North Carolina at

Charlotte, 9201 University City

Boulevard, North Carolina

Duke Energy Corporation, Docket Nos.

50-269, 50-270, and 50-287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of amendment request: October

20, 1997

Description of amendment request:

The proposed amendments would allow

use of a rerolling process as an

additional repair method for tube

degradation found in the tubesheet

region. The rerolling method is designed

to ensure that the area of degradation

will not serve as a pressure boundary

once the repair roll is installed, thus

permitting the tube to remain in service.

Date of publication of individual

notice in Federal Register: October 28,

1997 (62 FR 55835)

Expiration date of individual notice:

Comment period ends November 12,

1997; Notice period ends November 28,

1997

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina

Notice Of Issuance Of Amendments To

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

of 1954, as amended (the Act), and the

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.

Duquesne Light Company, et al., Docket

Nos. 50-334 and 50-412, Beaver Valley

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of application for amendments:

October 4, 1997

Brief description of amendments:

These amendments revise the

surveillance requirements in Technical

Specifications (TSs) 4.1.2.3.1, 4.1.2.4.1,

4.5.2.b, and 4.6.2.1.b and associated

Bases. The subject surveillance

requirements are applicable to the

charging/high-head safety injection

pumps, low-head safety injection

pumps, and the containment quench

spray pumps. The proposed changes

replace the current specific test

acceptance criteria contained in these

surveillance requirements with

requirements to verify pump

performance in accordance with the

inservice testing program, the

emergency core cooling system flow

analysis, or the containment integrity

safety analysis, as applicable. The

proposed changes also make minor

editorial changes in these TSs and make

conforming changes in the TS Index

pages.

Date of issuance: October 28, 1997

Effective date: Both units, as of date

of issuance, to be implemented within

60 days.

Amendment Nos.: 207, 86

Facility Operating License Nos. DPR66 and NPF-73: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: December 18, 1996 (61 FR

66706) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

October 28, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001

Entergy Operations, Inc., System

Energy Resources, Inc., South

Mississippi Electric Power Association,

and Mississippi Power & Light

Company, Docket No. 50-416, Grand

Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of application for amendment:

May 27, 1997, supplemented by October

6, 1997

Brief description of amendment: The

amendment eliminated selected

response time testing (RTT) surveillance

requirements (SRs) from the Technical

Specifications (TSs) for certain

components of the following systems:

reactor protection system (SR

3.3.1.1.15), primary containment and

drywell isolation instrumentation (SR

3.3.6.1.8), and emergency core cooling

system (SRs 3.5.1.8 and 3.5.2.7).

Date of issuance: November 5, 1997

Effective date: November 5, 1997

Amendment No.: 133

Facility Operating License No. NPF29: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: June 18, 1997 (62 FR 33122)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated November 5, 1997.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish,

Louisiana

Date of amendment request: July 17,

1996, as supplemented October 14, 1997

Brief description of amendment: The

amendment revises
61850 Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices

Facility Operating License No. NPF-38

to reflect the name change from

Louisiana Power & Light Company to

Entergy Louisiana, Inc.

Date of issuance: November 3, 1997

Effective date: November 3, 1997, to

be implemented within 60 days.

Amendment No.: 134

Facility Operating License No. NPF38: Amendment revised

Facility Operating License No. NPF38.

Date of initial notice in Federal

Register: April 9, 1997 (62 FR 29749)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated November 3, 1997.

The letter dated October 14, 1997,

provided clarifying information which

did not alter the initial no significant

hazards determination. No significant

hazards consideration comments

received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122

Florida Power Corporation, et al.,

Docket No. 50-302, Crystal River Unit

No. 3 Nuclear Generating Plant, Citrus

County, Florida

Date of application for amendment:

March 27, as supplemented April 3,

May 1, and August 20, 1997.

Brief description of amendment:

Change Technical Specifications (TS) to

permanently establish a primary-tosecondary leak rate of 150 gallons per

day through any one steam generator

and specify the steam generator tube

inservice inspection requirements for

pit-like intergranular attack degradation

in the ‘‘B’’ Once-Through-SteamGenerator.

Date of issuance: October 28, 1997

Effective date: October 28, 1997

Amendment No.: 158

Facility Operating License No. DPR72: Amendment revised the TS.

Date of initial notice in Federal

Register: June 4, 1997 (62 FR 30632)

The August 20, 1997, letter provided

clarifying information that did not affect

the initial no significant hazards

consideration. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

October 28, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida

32629

Florida Power and Light Company,

Docket Nos. 50-250 and 50-251, Turkey

Point Plant Units 3 and 4, Dade County,

Florida

Date of application for amendments:

August 27, 1997

Brief description of amendments: The

admendments change the

Administrative Section of the Technical

Specifications (TS) to allow the use of

12-hour shifts.

Date of issuance: October 27, 1997

Effective date: October 27, 1997

Amendment Nos: 194 and 188Facility

Operating Licenses Nos. DPR-31 and

DPR-41: Amendments revised the TS.

Date of initial notice in Federal

Register: September 24, 1997 (62 FR

50006) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

October 27, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199

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:

March 26, 1997.

Brief description of amendments: The

amendments modify surveillance

4.7.5.1.e.2 which requires verification of

the control room ventilation system

autostart function.

Date of issuance: October 28, 1997

Effective date: October 28, 1997, with

full implementation within 45 days.

Amendment Nos.: 218 and 202

Facility Operating License Nos. DPR58 and DPR-74: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: May 21, 1997 (62 FR 27796)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated October 28, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085

Northeast Nuclear Energy Company,

Docket No. 50-245, Millstone Nuclear

Power Station, Unit 1, New London

County, Connecticut

Date of application for amendment:

May 15, 1997

Brief description of amendment: The

amendment revises Technical

Specification Sections 3.1 and 4.1,

‘‘Reactor Protection System,’’ and the

associated Bases to remove run mode

intermediate range monitor high flux/

inoperative with the associated average

power range monitor downscale scram

trip function. The amendment also

makes other editorial revisions.

Date of issuance: October 27, 1997

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 103

Facility Operating License No. DPR21: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 18, 1997 (62 FR 33127)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated October 27, 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 at the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut 06385

Northeast Nuclear Energy Company, et

al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of application for amendment:

September 2, 1997

Brief description of amendment: The

amendment changes the Technical

Specifications (TSs) by modifying the

maximum allowed primary containment

internal pressure during normal

operation from 2.1 pounds per square

inch gauge (psig) to 1.0 psig. The TS

Bases, Section 3/4.6.1.4, is also updated

to reflect the new maximum allowed

primary containment internal pressure

during normal operation.

Date of issuance: October 27, 1997

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 209

Facility Operating License No. DPR65: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 24, 1997 (62 FR

50007) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

October 27, 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, and Waterford

Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, Connecticut
Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices 61851

Northern States Power Company,

Docket No. 50-263, Monticello Nuclear

Generating Plant, Wright County,

Minnesota

Date of application for amendment:

November 25, 1996, as supplemented

December 12, 1996, April 23, May 8,

July 1, August 21, and September 29,

1997

Brief description of amendment: The

amendment modifies the Technical

Specification requirements associated

with the Minimum Critical Power Ratio

(MCPR) safety limits for Cycle 18 based

on the cycle-specific analysis of the

current mixed core of GE11/GE10 fuel

parameters.

Date of issuance: October 29, 1997

Effective date: October 29, 1997

Amendment No.: 99

Facility Operating License No. DPR22: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 9, 1997 (62 FR 17238)

The December 12, 1996, letter provided

an affidavit for the original application

dated November 25, 1996. The April 23,

May 8, August 21, and September 29,

1997, letters provided clarifying

information in response to the staff’s

request for additional information

during a teleconference on March 18,

1997. The July 1, 1997, letter provided

a nonproprietary version of the April 23,

1997, submittal. This information was

within the scope of the original

application and did not change the

staff’s initial proposed no significant

hazards considerations determination.

Therefore, renoticing was not

warranted. The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

October 29, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room

location:Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Northern States Power Company,

Docket Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of application for amendments:

November 27, 1996, as supplemented

August 15, September 2, and October 3,

1997

Brief description of amendments: The

amendments incorporate Combustion

Engineering steam generator tube sleeve

designs and installation and

examination techniques into the plant

Technical Specifications (TS).

Specifically, the amendments make

changes to TS 4.12, ‘‘Steam Generator

Tube Surveillance,’’ and its associated

Bases Section B.4.12, ‘‘Steam Generator

Tube Surveillance.’’

Date of issuance: November 4, 1997

Effective date: November 4, 1997,

with full implementation within 30

days

Amendment Nos.: 132 and 124

Facility Operating License Nos. DPR42 and DPR-60: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: August 13, 1997 (62 FR 43370)

The August 15, September 2, and

October 3, 1997, letters provided

clarifying information and updated TS

pages. This information was within the

scope of the original application and did

not change the staff’s initial no

significant hazards considerations

determination. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

November 4, 1997.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

PECO Energy Company, Public Service

Electric and Gas Company Delmarva

Power and Light Company, and

Atlantic City Electric Company, Docket

Nos. 50-277 and 50-278, Peach Bottom

Atomic Power Station, Unit Nos. 2 and

3, York County, Pennsylvania

Date of application for amendments:

June 4, 1997

Brief description of amendments: The

proposed change revises the Peach

Bottom Atomic Power Station, Unit Nos.

2 and 3, Technical Specifications to

eliminate an inconsistency between

emergency core cooling system (ECCS)

operability requirements and the autostart and protective trip bypass of the

emergency diesel generators on an ECCS

initiation signal during certain plant

configurations.

Date of issuance: October 24, 1997

Effective date: Both units, as of date

of issuance, to be implemented within

30.

Amendments Nos.: 221 and 226

Facility Operating License Nos. DPR44 and DPR-56: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 13, 1997 (62 FR 43373)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated October 24, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, PA 17105

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

May 19, 1997, as supplemented by letter

dated August 25, 1997

Brief description of amendment: This

amendment changes the Hope Creek

Technical Specification (TS) 3.7.1.3,

‘‘Ultimate Heat Sink,’’ to raise the

minimum allowable ulimate heat sink

(UHS) water level from 76 feet to 80

feet, lower the maximum allowable UHS

temperature from 88.6°F to 85°F, and

reflect that continued plant operation to

a UHS temperature of 87°F depends

upon the association of UHS

temperature and safety system

availability. The associated Surveillance

Requirement, TS 4.7.1.3, is changed to

decrease the river water temperature, at

which increasing temperature

surveillance is required, from 85°F to

82°F. The requirements of TS 3.7.1.1,

‘‘Safety Auxiliaries Cooling System

(SACS),’’ TS 3.7.1.2, ‘‘Station Service

Water System (SSWS),’’ and TS 3.8.1.1,

‘‘Electrical Power Systems,’’ are revised

to reflect the revised TS 3.7.1.3. In

addition, the Bases for 3/4.7.1, ‘‘Service

Water Systems,’’ are appropriately

revised.

Date of issuance: October 28, 1997

Effective date: As of the date of

issuance to be implemented within 60

days from the date of issuance.

Amendment No.: 106

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: June 18, 1997 (62 FR 33132)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated October 28, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

March 31, 1997, as supplemented by

letters dated July 16, August 26, and

October 3, 1997

Brief description of amendment: This

amendment changes Technical

Specification (TS) 2.1.2, ‘‘THERMAL

POWER, High Pressure and High Flow,’’
61852 Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices

ACTION a.1.c for TS 3.4.1.1,

‘‘Recirculation Loops,’’ and the Bases

for TS 2.1, ‘‘Safety Limits.’’ These

changes are being made to implement

an appropriately conservative Safety

Limit Minimum Critical Power Ratio to

include Cycle 8 specific analyses for all

Hope Creek core and fuel designs.

Date of issuance: November 4, 1997

Effective date: The license

amendment is effective as of its date of

issuance and shall be implemented

within 60 days.

Amendment No.: 107

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 13, 1997 (62 FR 43374)

The August 26 and October 3, 1997,

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

November 4, 1997. No significant

hazards consideration comments

received: No.

Local Public Document Room

location:Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

September 29, 1997

Brief description of amendment: This

amendment changes Technical

Specification (TS) 3/4.11.1, ‘‘Liquid

Effluent - Concentration.’’ The change

adds a requirement to perform weekly

sampling and monthly and quarterly

composite analyses of the Station

Service Water System when the Reactor

Auxiliaries Cooling System is

contaminated. The licensee has also

proposed an editorial change to TS

Table 4.11.1.1.1-1. In Liquid Release

Type B, the licensee is proposing that

the acronym for Station Service Water

System be changed from GSW to SSWS.

This proposed change will be addressed

in a future license amendment.

Date of issuance: November 6, 1997

Effective date: As of date of issuance,

to be implemented within 60 days.

Amendment No.: 108

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: October 6, 1997 (62 FR 52161)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated November 6, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

September 24, 1997

Brief description of amendment: This

amendment adds a Surveillance

Requirement to Technical Specification

3/4.5.1, ‘‘Emergency Core Cooling

Systems’’, to perform a monthly valve

position verification for the four

residual heat removal cross-tie valves.

Date of issuance: November 6, 1997

Effective date: As of the date of

issuance, to be implemented within 60

days.

Amendment No.: 109

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: October 6, 1997 (62 FR 52162)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated November 6, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Public Service Electric & Gas Company,

Docket No. 50-354, Hope Creek

Generating Station, Salem County, New

Jersey

Date of application for amendment:

August 20, 1997

Brief description of amendment: This

amendment changes the Technical

Specifications (TSs) to provide for: 1)

the relocation of suppression chamber

volume references in Limiting

Condition for Operation (LCO) 3.5.3 to

the Hope Creek (HC) Updated Final

Safety Analysis Report (UFSAR) and TS

Bases as appropriate; 2) the revision of

the suppression chamber volume

currently listed in LCO 3.5.3.b; 3) the

relocation of the suppression chamber

volume references in LCO 3.6.2.1.a.1 to

the UFSAR and TS Bases; and 4) the

revision to the suppression chamber

volume reference in TS 5.2.1 to

reference the TS Bases section where

this information will reside.

Date of issuance: November 6, 1997

Effective date: As of date of issuance,

to be implemented within 60 days.

Amendment No.: 110

Facility Operating License No. NPF57: This amendment revised the

Technical Specifications and License.

Date of initial notice in Federal

Register: September 24, 1997 (62 FR

50010) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

November 6, 1997. No significant

hazards consideration comments

received: No.

Local Public Document Room

location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Rochester Gas and Electric

Corporation, Docket No. 50-244, R. E.

Ginna Nuclear Power Plant, Wayne

County, New York

Date of application for amendment:

August 19, 1997, as supplemented

September 29, 1997.

Brief description of amendment: The

proposed amendment revises the Ginna

Station Improved Technical

Specifications by adding a note to the

Containment Spray (CS) Limiting

Condition for Operation 3.6.6 which

allows the CS pumps in MODE 4 to be

placed in pull-stop, and motor-operated

valves 896A and 896B to have their DC

control power restored with the valves

placed in the closed position in order to

perform interlock and valve testing of

MOVs 857A, 857B, and 857C. A time

limit of 2 hours is placed on this

configuration for each test.

Date of issuance: October 29, 1997

Effective date: October 29, 1997

Amendment No.: 68

Facility Operating License No. DPR18: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 24, 1997 (62 FR

50011) The September 29, 1997, letter

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

October 29, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Rochester Public Library, 115

South Avenue, Rochester, New York

14610

Southern Nuclear Operating Company,

Inc., Docket Nos. 50-348 and 50-364,

Joseph M. Farley Nuclear Plant, Units

1 and 2, Houston County, Alabama

Date of amendments request:

September 17, 1997

Brief description of amendments: The

amendments change Technical

Specification 3/4.4.9, ‘‘Specific

Activity,’’ and the associated Bases to

reduce the limit associated with dose

equivalent iodine-131. The steady-state

dose equivalent iodine-131 limit would
Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices 61853

be reduced by 50 percent from 0.3 mu

Curie/gram to 0.15 mu Curie/gram and

the maximum instantaneous value

would be reduced by 50 percent from 18

mu Curie/gram to 9 mu Curie/gram.

Date of issuance: October 29, 1997

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: Unit 1 - 132; Unit

2 - 124

Facility Operating License Nos. NPF2 and NPF-8: Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: September 24, 1997 (62 FR

49998) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

October 29, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302

The Cleveland Electric Illuminating

Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison

Company, OES Nuclear, Inc.,

Pennsylvania Power Company, Toledo

Edison Company, Docket No. 50-440

Perry Nuclear Power Plant, Unit 1,

Lake County, Ohio

Date of application for amendment:

August 14, 1997, as supplemented

September 26 and October 1, 1997.

Brief description of amendment: This

amendment changes the design basis as

described in the Updated Safety

Analysis Report by adding a description

of the methodology utilized for

determining the systems and

components that are considered to

require protection from tornado

missiles.

Date of issuance: November 4, 1997

Effective date: November 4, 1997

Amendment No.: 90

Facility Operating License No. NPF58: This amendment revised the license.

Date of initial notice in Federal

Register: September 16, 1997 (62 FR

48674). The September 26 and October

1, 1997, submittals provided

supplemental 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

November 4, 1997. No significant

hazards consideration comments

received: No.

Local Public Document Room

location: Perry Public Library, 3753

Main Street, Perry, OH 44081

TU Electric Company, Docket Nos. 50-

445 and 50-446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2,

Somervell County, Texas

Date of amendment request: May 16,

1997 (TXX-97119)

Brief description of amendments: The

amendments revised core safety limit

curves and Overtemperature N-16

reactor trip setpoints based on analyses

of the core configuration for CPSES Unit

2, Cycle 4. These changes apply equally

to CPSES Units 1 and 2 licenses since

the Technical Specifications are

combined.

Date of issuance: October 30, 1997

Effective date: October 30, 1997

Amendment Nos.: Unit 1 -

Amendment No. 55; Unit 2 -

Amendment No. 41

Facility Operating License Nos. NPF87 and NPF-89: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: July 16, 1997 (62 FR 38140)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated October 30, 1997. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019

Notice Of Issuance Of Amendments To

Facility Operating Licenses And Final

Determination Of No Significant

Hazards Consideration And

Opportunity For A Hearing (Exigent

Public Announcement Or Emergency

Circumstances)

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application for the

amendment complies with the

standards and requirements of the

Atomic Energy Act 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.

Because of exigent or emergency

circumstances associated with the date

the amendment was needed, there was

not time for the Commission to publish,

for public comment before issuance, its

usual 30-day Notice of Consideration of

Issuance of Amendment, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for a

Hearing.

For exigent circumstances, the

Commission has either issued a Federal

Register notice providing opportunity

for public comment or has used local

media to provide notice to the public in

the area surrounding a licensee’s facility

of the licensee’s application and of the

Commission’s proposed determination

of no significant hazards consideration.

The Commission has provided a

reasonable opportunity for the public to

comment, using its best efforts to make

available to the public means of

communication for the public to

respond quickly, and in the case of

telephone comments, the comments

have been recorded or transcribed as

appropriate and the licensee has been

informed of the public comments.

In circumstances where failure to act

in a timely way would have resulted, for

example, in derating or shutdown of a

nuclear power plant or in prevention of

either resumption of operation or of

increase in power output up to the

plant’s licensed power level, the

Commission may not have had an

opportunity to provide for public

comment on its no significant hazards

consideration determination. In such

case, the license amendment has been

issued without opportunity for

comment. If there has been some time

for public comment but less than 30

days, the Commission may provide an

opportunity for public comment. If

comments have been requested, it is so

stated. In either event, the State has

been consulted by telephone whenever

possible.

Under its regulations, the Commission

may issue and make an amendment

immediately effective, notwithstanding

the pendency before it of a request for

a hearing from any person, in advance

of the holding and completion of any

required hearing, where it has

determined that no significant hazards

consideration is involved.

The Commission has applied the

standards of 10 CFR 50.92 and has made

a final determination that the

amendment involves no significant

hazards consideration. The basis for this

determination is contained in the

documents related to this action.

Accordingly, the amendments have

been issued and made effective 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
61854 Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices

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 application for

amendment, (2) the amendment to

Facility Operating License, 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 room for the

particular facility involved.

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By

December 19, 1997, the licensee may

file a request for a hearing with respect

to issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

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. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

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 the

factors specified in 10 CFR

2.714(a)(1)(i)-(v) and 2.714(d).

Boston Edison Company, Docket No.

50-293, Pilgrim Nuclear Power Station,

Plymouth County, Massachusetts

Date of application for amendment:

October 24, 1997

Brief description of amendment: The

amendment adds a footnote to

Technical Specification 3.7.A.5,

‘‘Primary Containment.’’ The footnote

provides a one time exception to the

reverse flow testing requirement for

containment isolation check valve 30-

CK-432.

Date of issuance: October 30, 1997

Effective date: As of date of issuance

and shall be implemented by November

2, 1997.

Amendment No.: 174

Facility Operating License No. DPR35: This amendment revised the

Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: No.

The Commission’s related evaluation

of the amendment, finding of emergency

circumstances, consultation with the

State of Massachusetts, and final no

significant hazards consideration

determination are contained in a Safety

Evaluation dated October 30, 1997.

Attorney for licensee: W. S. Stowe,

Esquire, Boston Edison Company, 800

Boylston Street, 36th Floor, Boston,

Massachusetts 02199

Local Public Document Room

location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360

NRC Project Director: Ronald B.

Eaton, Acting Director

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: October

17, 1997
Federal Register / Vol. 62, No. 223 / Wednesday, November 19, 1997 / Notices 61855

1The Manager is an investment adviser to the

Trust and serves in such capacity pursuant to a

contract subject to section 15 of the Act. All

registered open-end investment companies that

currently intend to rely on the order have been

named as applicants. Any other existing or future

investment company that relies on the order will

comply with the terms and conditions of the order.

Brief description of amendment: The

amendment revised Technical

Specification 4.5.2b and associated

Bases to eliminate the requirement to

vent the centrifugal charging pump

casings.

Date of issuance: November 3, 1997

Effective date: November 3, 1997

Amendment No.: 114

Facility Operating License No. NPF42: The amendment revised the

Technical Specifications. Press release

issued requesting comments as to

proposed no significant hazards

consideration: Yes. October 24, 1997.

Coffey County Today Newspaper

(Kansas). Comments received: Yes.

Comments were submitted by Mr. Dave

Lochbaum of the Union of Concerned

Scientists by letter dated October 29,

1997. Verbal comments were received

from Larry Myers on October 28, 1997.

The staff responded to these comments

in the safety evaluation attached to the

November 3, 1997, amendment. The

Commission’s related evaluation of the

amendment, finding of exigent

circumstances, consultation with the

State of Kansas and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated November 3, 1997.

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts and Trowbridge,

2300 N Street, N.W., Washington, D.C.

20037

NRC Project Director: William H.

Bateman

Dated at Rockville, Maryland, this

12th day of November 1997.

For the Nuclear Regulatory Commission

Elinor G. Adensam,

Acting Director, Division of Reactor Projects

- III/IV Office of Nuclear Reactor Regulation.

[FR Doc. 97–30217 Filed 11–18–97; 8:45 am]

BILLING CODE 7590–01–F

SECURITIES AND EXCHANGE

COMMISSION

[Rel. No. IC–22883/812–10536]

EQ Advisors Trust; Notice of

Application

November 12, 1997.

AGENCY: Securities and Exchange

Commission (‘‘SEC’’).

ACTION: Notice of Application for an

order under (i) section 6(c) of the

Investment Company Act of 1940 (the

‘‘Act’’) granting relief from sections

13(a)(2), 18(f)(1), 22(f), and 22(g) of the

Act; and (ii) section 17(d) of the Act and

rule 17d–1 to permit certain joint

transactions.

Summary of Application: Applicants

request an order to permit EQ Advisors

Trust to implement a deferred

compensation plan for certain of its

trustees.

Applicants: HQ Advisors Trust (the

‘‘Trust’’) and EQ Financial Consultants,

Inc. (the ‘‘Manager’’).

FILING DATES: The application was

filed on April 7, 1997, and amendments

were filed on July 14, 1997 and

November 10, 1997.

Hearing or Notification of Hearing: An

order granting the application will be

issued unless the SEC orders a hearing.

Interested persons may request a

hearing by writing to the SEC’s

Secretary and serving applicant with a

copy of the request, personally or by

mail. Hearing requests should be

received by the SEC by 5:30 p.m. on

December 8, 1997, and should be

accompanied by proof of service on

applicant, in the form of an affidavit or,

for lawyers, a certificate of service.

Hearing requests should state the nature

of the writer’s interest, the reason for the

request, and the issues contested.

Persons who wish to be notified of a

hearing may request notification by

writing to the SEC’s Secretary.

ADDRESSES: Secretary, SEC, 450 Fifth

Street, N.W., Washington, D.C. 20549.

EQ Advisors Trust, 1290 Avenue of the

Americas, New York, New York 10104.

FOR FURTHER INFORMATION CONTACT:

Deepak T. Pai, Staff Attorney, at (202)

942–0574, or Nadya B. Roytblat,

Assistant Director, at (202) 942–0564

(Division of Investment Management,

Office of Investment Company

Regulation).

SUPPLEMENTARY INFORMATION: The

following is a summary of the

application. The complete application

may be obtained for a fee at the SEC’s

Public Reference Branch, 450 Fifth

Street, N.W., Washington, D.C. 20549

(tel. 202–942–8090).

Applicants’ Representations

1. The Trust is an open-end

management investment company

registered under the Act and organized

as a Delaware business trust. The Trust

is currently composed of several,

separately managed series

(‘‘Portfolios’’). The Trust offers shares in

each of its Portfolios only to insurance

companies and their separate accounts

that fund variable annuity and variable

life insurance contracts (‘‘Variable

Contracts’’). The Trust is currently

serving as the underlying investment

medium for Variable Contracts issued

by the Equitable Life Assurance Society

of the United States (‘‘Equitable’’). The

Trust may in the future offer its shares

to separate accounts funding Variable

Contracts of insurance companies

unaffiliated with Equitable or directly to

tax qualified pension and retirement

plans outside the separate account

context.

2. The Manager, an indirect whollyowned subsidiary of Equitable, has

overall responsibility for the investment

management and administration of the

Trust and its Portfolios. Rowe PriceFleming International, Inc., T. Rowe

Price Associates, Inc., Putnam

Investment Management, Inc.,

Massachusetts Financial Services

Company, Morgan Stanley Asset

Management, Inc., Warburg Pincus

Counsellors, Inc., and Merrill Lynch

Asset Management, L.P. serve as the

sub-advisers (each an ‘‘Adviser’’) to one

or more Portfolios. Applicants request

that the relief apply to the Trust and any

registered open-end management

investment company that in the future

is advised by the Manager or any entity

controlled by the Manager.1

3. The Trust’s board of trustees

(‘‘Trustees’’) currently consists of six

members, two of whom are ‘‘interested

persons’’ of the Trust within the

meaning of Section 2(a)(19) of the Act.

The four non-interested Trustees

(‘‘Eligible Trustees’’) will receive an

annual retainer fee, a fee for each board

meeting and committee meeting

attended, and an additional fee for

performing special services for the

Trust.

4. The deferred compensation plan for

Eligible Trustees (the ‘‘Plan’’) was

ratified by the Trustees on March 31,

1997. The purpose of the Plan is to

permit Eligible Trustees to defer receipt

of all or a portion of their fees to enable

them to defer payment of income taxes,

to avoid a loss or reduction of Social

Security benefits, or for other reasons.

Applicants believe that the Plan will

better enable the Trust to attract and

retain high caliber trustees. The Plan

may be amended from time to time,

provided that, any amendments are not

inconsistent with the relief granted

pursuant to this application.

5. Under the Plan, each Eligible

Trustee who elects to defer receipt of