Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices 11483

from simplified operations and the

reduced potential for undesirable

opening of ERV will more than offset

the reduction of the principal membrane

safety factor. Reduced operational

challenges will reduce the potential for

undesirable impacts to the environment.

The change will not increase the

probability or consequences of

accidents, no changes are being made in

the types of any effluents that may be

released offsite, and there is no

significant increase in the allowable

individual or cumulative occupational

radiation exposure. Accordingly, the

Commission concludes that there are no

significant radiological environmental

impacts associated with the proposed

action.

With regard to potential

nonradiological impacts, the proposed

action involves features located entirely

within the restricted area as defined in

10 CFR Part 20. It does not affect

nonradiological plant effluents and has

no other environmental impact.

Accordingly, the Commission concludes

that there are no significant

nonradiological environmental impacts

associated with the proposed action.

Alternatives to the Proposed Action

Since the Commission has concluded

there is no measurable environmental

impact associated with the proposed

action, any alternatives with equal or

greater environmental impact need not

be evaluated. As an alternative to the

proposed action, the staff considered

denial of the proposed action. Denial of

the application would result in no

change in current environmental

impacts. The environmental impacts of

the proposed action and the alternative

action are similar.

Alternative Use of Resources

This action does not involve the use

of any resources not previously

considered in the Final Environmental

Statement for ANO–1.

Agencies and Persons Consulted

In accordance with its stated policy,

on January 28, 1996, the staff consulted

with the Arkansas State official, Mr.

David Snellings, Director of the Division

of Radiation Control and Emergency

Management, regarding the

environmental impact of the proposed

action. The State official had no

comments.

Finding of No Significant Impact

Based upon the environmental

assessment, the Commission concludes

that the proposed action will not have

a significant effect on the quality of the

human environment. Accordingly, the

Commission has determined not to

prepare an environmental impact

statement for the proposed action.

For further details with respect to the

proposed action, see the licensee’s letter

dated November 26, 1996, which is

available for public inspection at the

Commission’s Public Document Room,

2120 L Street, NW., Washington, DC,

and at the local public document room

located at the Tomlinson Library,

Arkansas Tech University, Russellville,

AR 72801.

Dated at Rockville, Maryland, this 7th day

of March 1997.

For the Nuclear Regulatory Commission.

George Kalman,

Senior Project Manager, Project Directorate

VI–1, Division of Reactor Projects III/IV, Office

of Nuclear Reactor Regulation.

[FR Doc. 97–6342 Filed 3–11–97; 8:45 am]

BILLING CODE 7590–01–P

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

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 February 14,

1997, through February 28, 1997. The

last biweekly notice was published on

February 26, 1997.

Notice of Consideration of Issuance of

Amendments to Facility Opeating

Licenses, Proposed No Significant

Harzards Consideration determination,

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

no significant hazards consideration.

Under the Commission’s regulations in

10 CFR 50.92, this means that operation

of the facility in accordance with the

proposed amendment would not (1)

involve a significant increase in the

probability or consequences of an

accident previously evaluated; or (2)

create the possibility of a new or

different kind of accident from any

accident previously evaluated; or (3)

involve a significant reduction in a

margin of safety. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

within 30 days after the date of

publication of this notice will be

considered in making any final

determination.

Normally, the Commission will not

issue the amendment until the

expiration of the 30-day notice period.

However, should circumstances change

during the notice period such that

failure to act in a timely way would

result, for example, in derating or

shutdown of the facility, the

Commission may issue the license

amendment before the expiration of the

30-day notice period, provided that its

final determination is that the

amendment involves no significant

hazards consideration. The final

determination will consider all public

and State comments received before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after issuance. The Commission

expects that the need to take this action

will occur very infrequently.

Written comments may be submitted

by mail to the Chief, Rules Review 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 April 11, 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
11484 Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices

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:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

Union at 1–(800) 248–5100 (in Missouri

1–(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–0001, and to the attorney for

the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

factors specified in 10 CFR 2.714(a)(1)

(i)–(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendment which is available for

public inspection at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC, and at the local public

document room for the particular

facility involved.

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant, Units 1

and 2, Brunswick County, North

Carolina

Date of amendment requests:

December 4, 1996

Description of amendments request:

The proposed amendments would

revise the Technical Specifications (TS)

to reflect a change in the method for

detecting a reactivity anomaly described

in TS 3.1.2 and TS Surveillance

Requirement 4.1.2. Actual keff will be

compared to predicted core keff instead

of comparing actual and predicted

control rod density to determine if a

reactivity anomaly exists. Additionally,

editorial changes to the Bases for TS 3/

4.1.2 are proposed to support the TS

amendments.

Basis for proposed no significant

hazards consideration determination: As

required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendments do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. The proposed license

amendments modify the method of detecting

a reactivity anomaly. The proposed license

amendments allow using core keff to detect a

reactivity anomaly instead of control rod

density. The correlation between core
Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices 11485

reactivity and control rod density depends on

predicting core keff. Core keff can be readily

monitored with the new plant process

computer program and core keff can more

accurately detect a reactivity anomaly in the

core (assumptions are minimized). A

reactivity anomaly is not considered an

initiator of any previously analyzed accident.

As such, changing the method of detecting a

reactivity anomaly will not increase the

probability of any accident previously

evaluated. Although, a reactivity anomaly

could impact the consequences of a

previously analyzed accident, the

consequences of an event occurring using the

proposed method of detecting a reactivity

anomaly are the same as the consequences of

an event occurring using the current method

of detecting a reactivity anomaly. As a result,

the proposed amendments do not involve a

significant increase in the consequences of

any accident previously evaluated.

2. The proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated. The proposed license

amendments do not involve a physical

modification to the plant. The proposed

license amendments also continue to verify

that the reactivity difference between

predicted and actual are such that a reactivity

anomaly does not exist. In addition, core keff

can more accurately detect a reactivity

anomaly in the core (assumptions are

minimized) and can be readily monitored

with the new plant process computer

program. Therefore, the change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed license amendments do

not involve a significant reduction in a

margin of safety. The proposed license

amendments modify the method of detecting

a reactivity anomaly. The proposed license

amendments allow using core keff to detect a

reactivity anomaly instead of control rod

density. The correlation between core

reactivity and control rod density depends on

predicting core keff. Core keff can be readily

monitored with the new plant process

computer, and core keff can more accurately

detect a reactivity anomaly in the core

(assumptions are minimized). Therefore, the

proposed license amendments 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: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297.

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: Mark Reinhart,

Acting

Carolina Power & Light Company, et al.,

Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant (BSEP),

Units 1 and 2, Brunswick County, North

Carolina

Date of amendment requests: January

7, 1997.

Description of amendments request:

The proposed amendments would

revise the Technical Specifications (TS)

to: (1) exchange the reactor pressure

vessel pressure-temperature (P–T) limits

curves currently located in the Unit 1

and 2 TS; and (2) delete the current 8,

10, and 12 effective full power year

(EFPY) hydrostatic test P–T limits

curves and incorporate new 14 and 16

EFPY hydrostatic test P–T limits curves

for the Unit 1 and 2 reactor pressure

vessels. As reported in Licensee Event

Report (LER) 1–94–05 dated March 22,

1994, and LER supplements dated April

29, 1994, and September 23, 1994, the

licensee, the Carolina Power & Light Co.

(CP&L), determined that the Unit 1 and

2 P–T limits curves had been

inadvertently transposed and evaluated

the effects of the transposition. The

proposed amendments correct this

transposition error. The proposed

changes to the hydrostatic test P–T

limits curves are required because it is

anticipated that both units will exceed

12 EFPY during 1997.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

This Technical Specification Change

Request makes the following changes:

1. Exchanges the pressure-temperature

limits curves currently located in the Unit 1

and Unit 2 Technical Specifications. In

Licensee Event Report 1–94–05, CP&L

reported that the Unit 1 and Unit 2 pressuretemperature limits curves had been

inadvertently transposed. This request is an

administrative change to relocate the

pressure-temperature limits curves to

Technical Specifications of the unit to which

they correctly correspond.

2. Deletes the current 8, 10 and 12 effective

full power year (EFPY) hydrostatic test

pressure-temperature limits curves and

incorporates new 14 and 16 effective full

power year (EFPY) hydrostatic test pressuretemperature limits curves for the Brunswick

Unit 1 and 2 reactors. The current reactor

vessel pressure-temperature limits curves

contained in the technical specifications for

hydrostatic pressure tests are suitable for up

to 12 effective full power years (EFPY) of

reactor operation. It is anticipated that both

units will surpass this threshold during 1997.

Based on this, new pressure-temperature

limits curves for 14 and 16 EFPY were

developed. Commensurate changes to the

references in Technical Specification 3/

4.4.6.1 and Bases 3/4.6 are also proposed to

reflect the deletion of current Technical

Specification Figure 3.4.6.1–3c.

3. Reformat[s] the pressure-temperature

limits curves in Technical Specification

Figures 3.4.6.1–1, 3.4.6.1–2, 3.4.6.1–3a, and

3.4.6.1–3b. The changes associated with

reformatting the Figures are administrative in

nature.

Items 1, 2, and 3 do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because of the following reasons:

1. Item 1 will exchange the Unit 1 and Unit

2 pressure-temperature limits curves. This

change is considered administrative in

nature. The pressure-temperature limits

curves were developed based on design and

materials information for the reactor vessel;

however, due to an administrative error

during the development of the curves, the

materials information for the Unit 1 and Unit

2 reactor vessels was inadvertently reversed.

Proposed change 1 is being made to exchange

the reactor coolant system pressuretemperature limits curves. Therefore, since

this proposed change does not involve a

change to the pressure-temperature limits

curves nor a change to the configuration of

the facility, the probability of an accident

previously evaluated is not increased.

Item 2 deletes the current Technical

Specification hydrostatic test pressuretemperature limits curves and replaces them

with updated curves. The current hydrostatic

test pressure-temperature limits curves,

which are valid through 12 EFPY are

expected to expire during 1997; therefore,

new hydrostatic test pressure-temperature

limits curves were developed through 16

EFPY. These new hydrostatic test pressuretemperature limits curves will ensure that the

integrity of the Brunswick Units 1 and 2

reactor pressure vessels is maintained during

hydrostatic and leak tests up to 16 effective

full power years of operation. The

calculations used to generate the new

pressure-temperature limits curves were

performed using Appendix G to Section XI of

the ASME Boiler and Pressure Vessel Code,

Welding Research Council Bulletin 175, and

Appendix A to Section XI of the ASME

Boiler and Pressure Vessel Code, and

[incorporate] the requirements of 10 CFR 50,

Appendix G, Section IV.A.2. For pressuretemperature limit curve development, the

methods described in Appendix G to Section

XI of the ASME Boiler and Pressure Vessel

Code are equivalent to the methods described

in Appendix G to Section III of the ASME

Boiler and Pressure Vessel Code. The

proposed pressure-temperature limits curves,

for hydrostatic and leak tests, take into

consideration the effects of neutron

irradiation on reactor vessel materials and

provide the necessary margin, as specified by

Appendix G of 10 CFR 50, to assure the

structural integrity of the reactor coolant

pressure boundary. Based on the above, it is

concluded that this change will not increase

the probability of an accident previously

evaluated.
11486 Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices

Item 3 reformats each of the Technical

Specification Figures containing the

pressure-temperature limits curves. The

changes associated with the reformatting of

proposed Technical Specification Figures

3.4.6.1–1, 3.4.6.1–2, 3.4.6.1–3a, and 3.4.6.1–

3b reflect presentation preferences and do

not result in technical changes (either actual

or interpretational) to the requirements of the

pressure-temperature limits curves.

Therefore, the changes associated with

reformatting the Technical Specification

Figures containing the pressure-temperature

limits curves are considered to be

administrative in nature. Based on the above,

it is concluded that this change will not

increase the probability of an accident

previously evaluated.

The proposed license amendments do not

alter Limiting Safety System Settings nor

Safety Limits. The proposed license

amendments do not revise the technical

bases from which the pressure-temperature

limits curves were derived, and do not affect

stresses and fatigue for transients and design

basis events for which the reactor vessels

were designed. The operation of plant

equipment is not significantly impacted by

the proposed license amendments. The

proposed pressure-temperature limits curves

provide the necessary margin to ... assure the

structural integrity of the reactor coolant

pressure boundary is maintained. This

margin is designed to preclude the

probability of a reactor coolant pressure

boundary failure. In addition, since the

proposed pressure-temperature limits curves

are based on current regulatory requirements

and fluence data, the consequences of a

reactor coolant pressure boundary failure are

not impacted by the proposed license

amendments. Therefore, the proposed license

amendments do not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. The proposed license amendments will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated. The proposed license

amendments will ensure that acceptable

pressure-temperature limits are imposed on

the reactor pressure vessels during all phases

of plant operation, thereby ensuring the

structural integrity of the reactor pressure

vessels. The pressure-temperature limits

curves are designed to provide fracture

protection for the reactor coolant pressure

boundary and do not create any new accident

modes. Accident modes for the reactor

coolant pressure boundary, due to nonductile

failure, are well understood by the industry.

The proposed pressure-temperature limits

curves and the Technical Specifications

continue to provide controls to preclude such

a failure. In addition, the proposed license

amendments do not result in physical

changes to the facility, nor do the proposed

license amendments alter safety-related

equipment, or safety functions. Therefore, the

proposed license amendments do not create

a new or different kind of accident from any

previously evaluated.

3. The proposed license amendments do

not involve a significant reduction in a

margin of safety. The pressure-temperature

limits curves are designed to provide a

specific margin of safety. This margin is

required to be at least as great as that

specified in Appendix G to Section III of the

ASME Boiler and Pressure Vessel Code and

Appendix G to 10 CFR 50. The proposed

pressure-temperature limits curves were

developed based on design and materials

information for the reactor vessels, current

regulatory requirements and fluence data.

The proposed pressure-temperature limit

curves are based on analyses that ensure that

the fracture toughness margins of 10 CFR Part

50, Appendix G are not exceeded. Therefore,

the proposed license amendments do not

involve a significant reduction in the margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297.

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: Mark Reinhart,

Acting.

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

455, Byron Station, Unit Nos. 1 and 2,

Ogle County, Illinois Docket Nos. STN

50–456 and STN 50–457, Braidwood

Station, Unit Nos. 1 and 2, Will County,

Illinois

Date of amendment request: April 29,

1996, as supplemented on January 21,

1997.

Description of amendment request:

The proposed amendment would:

1. Revise Technical Specification (TS)

3.7.1.1, Action a., to require the unit to

be in hot shutdown, rather than cold

shutdown, for consistency with

NUREG–1431, ‘‘Standard Technical

Specifications for Westinghouse

Plants,’’ and add a new Action b. to

clarify the shutdown requirements

when there are more than three

inoperable main steam line Code safety

valves on any one steam generator.

2. Revise TS Surveillance

Requirement 4.7.1.1 to clarify that

Specification 4.0.4 does not apply for

entry into Mode 3 for Byron and

Braidwood and, for Braidwood only,

delete the one-time requirements for

Unit 1, Cycle 5 and Unit 2 after outage

A2F27.

3. Revise the maximum allowable

power range neutron flux high trip

setpoints in Table 3.7–1.

4. Revise Table 3.7–2 to increase the

as-found main steam safety valve

(MSSV) lift setpoint tolerance to plus/

minus 3%, provide an as-left setpoint

tolerance of plus/minus 1%, and change

a table notation.

5. Delete the orifice size column from

Table 3.7–2.

6. Revise the Bases for TS 3.7.1.1 to

be consistent with the proposed changes

to TS 3.7.1.1.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The text describing reactor coolant loops

and steam generators is redundant. TS

3.4.1.1, ‘‘Reactor Coolant Loops and Coolant

Circulation—Startup and Power Operation,’’

and 3.4.1.2, ‘‘Reactor Coolant Loops and

Coolant Circulation—Hot Standby,’’ provide

restrictions on the number of operating

reactor coolant loops and steam generators.

Therefore, deleting the text that requires

having four reactor coolant loops and

associated steam generators in operation from

TS 3.7.1.1, Action a., has no impact on any

analyzed accident.

The proposed change to TS 3.7.1.1, Action

a., to require the final mode to be hot

shutdown rather than cold shutdown is

consistent with the Applicability section of

the specification, which does not require the

MSSVs to be operable in hot shutdown.

There are no credible transients requiring the

MSSVs in modes 4 and 5. The steam

generators are not normally used for heat

removal in modes 5 and 6, and thus cannot

be overpressurized. The change also

eliminates the unnecessary transient that had

been imposed on the unit by forcing entry

into cold shutdown.

The new Action b. for TS 3.7.1.1 and text

changes to Action a. clarify the shutdown

requirement times based on the number of

inoperable valves. There are no changes to

these times.

Changing TSSR 4.7.1.1 to delete the onetime requirements imposed by previous

amendments and allow entry into Mode 3

prior to performing the requirements of TSSR

4.0.5 has no impact on any accident. The

change permits testing the MSSVs in

accordance with the applicable codes and

allows a reasonable amount of time for

completion of the surveillance. The

conditions requiring the one-time

requirements have been corrected, so the

one-time requirements are no longer

required.

The proposed setpoints in Table 3.7–1 are

more limiting than those currently allowed in

Specification 3.7.1.1. Westinghouse
Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices 11487

determined that the current setpoints are

non-conservative for some combinations of

reduced MSSV availability and reactor power

levels. By reducing the setpoints, the original

design margins for safety will be met.

Reduced reactor trip setpoints due to reduced

availability of the MSSVs are not precursors

to any accidents, but are used in the safety

analysis to establish that plant response will

be within required margins for accidents of

concern.

Increasing the as-found valve setpoint

tolerance from plus/minus 1% to plus/minus

3% does not have a significant impact on any

accident. The peak primary and secondary

pressures remain below 110% of design at all

times. The departure from nucleate boiling

ratio and peak cladding temperature values

remain within the specified limits of the

licensing basis. All of the applicable loss-ofcoolant accident (LOCA) and non-LOCA

design basis acceptance criteria remain valid.

The MSSVs are actuated after accident

initiation to protect the secondary systems

from overpressurization. Increasing the asfound setpoint tolerance will not result in

any hardware modification to the MSSVs.

Therefore, there is not an increase in the

probability of the spurious opening of a

MSSV. Sufficient margin exists between the

normal steam system operating pressure and

the valve setpoint with the increased

tolerance to preclude an increase in the

probability of actuating the valves. The

MSSVs also remain capable of relieving any

unlikely system overpressure during all

applicable operating modes.

Although increasing the as-found valve

setpoint tolerance may increase the steam

release from the ruptured steam generator

above the Updated Final Safety Analysis

Review (UFSAR) value by approximately 2%,

the steam generator tube rupture analysis

indicates that the calculated break flow is

still less than the value reported in the

UFSAR. Therefore, the radiological analysis

indicates that the slight increase in the steam

release is offset by the decrease in the break

flow such that the offsite radiation doses are

less than those reported in the UFSAR. The

evaluation also concluded that the existing

mass releases used in the offsite dose

calculation for the remaining transients (i.e.,

steam line break, rod ejection) are still

applicable. Therefore, based on the above,

there is no increase in the dose releases.

Neither the mass and energy release to the

containment following a postulated LOCA,

nor the analysis of containment response

following the LOCA credit the MSSVs in

mitigating the consequences of an accident.

Therefore, changing the MSSV lift setpoint

tolerances would have no impact on the

containment integrity analysis. In addition,

based on the conclusion of the transient

analysis, the change to the MSSV tolerance

will not affect the calculated steam line break

mass and energy releases inside containment.

Deleting the orifice size column from Table

3.7.1–2 has no impact on previously

evaluated accidents. There is no change to

the orifice size, which is stated in the UFSAR

and incorporated as needed in the accident

analyses.

The proposed changes do not introduce

any new equipment, equipment

modifications, or any new or different modes

of plant operation. The MSSVs are not

precursors to any analyzed accident. The

proposed changes will not affect the

operational characteristics of any equipment

or systems.

Therefore, these proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

B. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Deleting the text describing reactor coolant

loops and steam generators from TS 3.7.1.1

Action a. has no impact on plant operation

since the specific restrictions on the number

of operating reactor coolant loops and steam

generators are provided in TS 3.4.1.1 and

3.4.1.2.

The proposed change to TS 3.7.1.1, Action

a., to require the final mode to be hot

shutdown rather than cold shutdown is

consistent with the Applicability section of

the specification, which does not require the

MSSVs to be operable in hot shutdown.

There are no credible transients requiring the

MSSVs in Modes 4 and 5. The steam

generators are not normally used for heat

removal in Modes 5 and 6, and thus cannot

be overpressurized. NUREG–1431 does not

include requirements for the MSSVs to be

operable in these modes. The change will

also eliminate the unnecessary transient that

had been imposed on the unit by forcing

entry into cold shutdown.

The new Action b. for TS 3.7.1.1 and text

changes to Action a. clarify the shutdown

requirement times based on the number of

inoperable valves. There are no changes to

the times.

The proposed change to TSSR 4.7.1.1 to

clarify that TSSR 4.0.4 does not apply for

entry into Mode 3 will allow ComEd to

continue to perform MSSV testing at normal

operating pressure and temperature as

required by the applicable codes. The change

precludes having to enter an action statement

to perform the testing and eliminates severe

time restrictions on the valve testing and

conflicts with other plant startup

requirements.

The proposed recalculated setpoints of

Table 3.7–1 are more limiting than those

currently allowed in the Specification and

ensure that the original design margins for

safety are met. The secondary system

pressure remains within design limits.

Increasing the as-found tolerance on the

MSSV setpoint to plus/minus 3% will not

increase the challenge to the MSSVs or result

in increased actuation of the valves. The

changes to the Bases document the method

for calculating the reduced reactor trip

setpoints based on reduced availability of

MSSVs.

Deleting the orifice size column from Table

3.7–2 and the obsolete one-time requirements

in TSSR 4.7.1.1 are administrative changes

only.

Increasing the lift setpoint tolerance on the

MSSVs does not introduce a new accident

initiator mechanism. The proposed change

does not introduce any new equipment,

equipment modifications, or any new or

different modes of plant operation. No new

failure modes have been defined for any

system or component important to safety nor

has any new limiting single failure been

identified. This change will not affect the

operational characteristics of any equipment

or systems. Thus, there is no change in the

margin for safety.

Therefore, these proposed changes will not

create the possibility of a new or different

type of accident from any accident

previously evaluated.

C. The proposed change does not involve

a significant reduction in a margin of safety.

Deleting the text describing reactor coolant

loops and steam generators has no impact on

plant operation since the specific restrictions

on the number of operating reactor coolant

loops and steam generators are provided in

TS 3.4.1.1 and 3.4.1.2.

The change requiring hot shutdown

instead of cold shutdown entry is more

appropriate than the existing specification

since the action statement places the plant in

a mode where operability of the MSSVs is

not required. The Technical Specification is

applicable in Modes 1, 2, and 3, therefore,

entering Mode 4 places the plant in a

condition where the MSSVs are not required

to be operable. There are no credible

transients requiring the MSSVs in Modes 4

and 5. The steam generators are not normally

used for heat removal in Modes 5 and 6, and

thus cannot be overpressurized. NUREG–

1431 does not include requirements for the

MSSVs to be operable in these modes.

Changing the mode in which the MSSVs

are tested will not change the operational

characteristics of the MSSVs. ComEd will

continue to test the MSSVs at normal

operating pressure and temperature as

required by the applicable codes.

The proposed reactor trip setpoints in

Table 3.7–1 are more limiting than the

current setpoints in the Specification.

Reactor trip settings were calculated using a

revised methodology to account for the nonlinear relationship of reactor trip setpoints

and reduced MSSV availability. The revised

setpoints ensure the original design margin of

safety is maintained. The proposed changes

to the Bases include the revised equation

used to calculate the reduced reactor trip

setpoints.

Increasing the as-found lift setpoint

tolerance on the MSSVs will not adversely

affect the operation of the reactor protection

system, any of the protection setpoints, or

any other device required for accident

mitigation. The proposed increase in the

setpoint tolerance does not invalidate the

LOCA and non-LOCA conclusions presented

in the UFSAR accident analyses. In letter

CAE–91–209/CAE 91–219, Westinghouse

concluded that the new loss of load/turbine

trip analysis satisfied all applicable

acceptance criteria and demonstrated that the

conclusion presented in the UFSAR remains

valid. For all the UFSAR non-LOCA

transients, the departure from nucleate

boiling design basis, primary and secondary

pressure limits, and dose release limits

continue to be met. Peak cladding

temperatures remain well below the limits

specified in the 10 CFR 50.46.
11488 Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices

Deleting the orifice size column from Table

3.7–2 and the obsolete one-time requirements

in TSSR 4.7.1.1 are administrative changes.

The proposed changes do not introduce

any new equipment, equipment

modifications, or any new or different modes

of plant operation. These changes will not

affect the operational characteristics of any

equipment or systems. Therefore, no

reduction in the margin of safety will occur

as a result of changes.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O.

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

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. STN 50–454 and STN 50–

455, Byron Station, Unit Nos. 1 and 2,

Ogle County, Illinois

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1 and

2, Will County, Illinois

Date of amendment request: August

23, 1996.

Description of amendment request:

The proposed amendment would revise

the technical specifications to reflect the

design lineup for the Non-Accessible

Area Exhaust Filter Plenum Ventilation

System, and to make provisions for the

performance of maintenance and testing

on the system.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

Non-Accessible Area Exhaust Filter

Plenum Ventilation (VA) System lineups are

not considered as the precursors to any

accident. The additional provisions added to

the action statement for TS 3.7.7

accommodates required maintenance and

surveillance activities. No new equipment is

being installed and no existing equipment is

being modified. Thus, these proposed

changes will not result in an increase in the

probability of occurrence of an accident

previously evaluated.

On the postulated Loss Of Coolant

Accident (LOCA) with Loss Of Offsite Power

(LOOP), the operating plenum will either

realign immediately or following the reenergization of its ESF bus which will occur

within 10 seconds. Thus, there will always

be at least one plenum operating immediately

during an accident. The emergency

procedures direct the realignment of the

standby plenum. This direction is contained

in the Byron and Braidwood Emergency

Procedures (BEP/BwEP)–0, ‘‘Reactor Trip or

Safety Injection,’’ and is performed prior to

conducting event diagnostic steps.

Filtration of the air from the Emergency

Core Cooling System (ECCS) equipment

cubicles becomes critical when the ECCS

pumps begin pumping accident water from

the containment recirculation sumps. Prior to

this the water flowing in these pumps is

Refueling Water Storage Tank (RWST) water.

This swap over from the RWST to the

containment recirculation sump is expected

to occur, at the earliest, 11 minutes following

accident initiation leaving time to open the

inlet damper on the standby VA plenum.

Thus, since the standby plenum can be

realigned before filtration of the ECCS

equipment cubicle air is required, the

Updated Final Safety Analysis Report

(UFSAR) assumptions, and offsite dose

calculation assumptions remain valid. There

will be no significant change in the types or

significant increase in the amounts of any

effluent that may be released offsite, and

there will be no significant increase in

individual or cumulative occupational

radiation exposure. Observations conducted

on licensed operators undergoing simulator

training verified that the VA system is

realigned well before the swap-over to the

containment recirculation sump under these

conditions. Therefore, these proposed

changes will not result in a significant

increase in the consequences of an accident

previously evaluated.

A review of the Byron and Braidwood

Probabilistic Risk Assessment (PRA) shows

that these proposed changes will have no

effect on either Core Damage Frequency

(CDF) or Uncontrolled Release Frequency

(URF).

Therefore, these changes do not involve a

significant increase in the probability or

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.

These proposed changes continue to

ensure that, following a LOCA, the air being

exhausted from the ECCS equipment rooms

is properly filtered before being released to

the environment.

These changes will not result in the

installation of any new equipment or the

modification of any existing equipment. No

new operating modes or system interfaces

will be created. The VA system will continue

to operate as designed during normal and

post accident conditions. All of the accident

analysis assumptions and conditions will

remain satisfied.

Thus this proposed change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

These proposed changes reflect the design

lineup for the VA system and provide action

requirements to accommodate required

maintenance and surveillance testing. The

VA system will continue to ensure that

following a LOCA, the air being exhausted

from the ECCS equipment rooms is properly

filtered before being released to the

environment.

Filtration of the ECCS equipment cubicle

air does not become critical until the suction

of the ECCS pumps is switched from the

RWST to the containment recirculation

sumps. This is postulated to occur, at the

earliest, 11 minutes following accident

initiation. On the postulated LOCA with

LOOP, at lease one VA plenum will be in

operation immediately and the emergency

procedures direct the realignment of the

standby plenum well before the ECCS pump

suction swap-over. Observations conducted

on licensed operators undergoing simulator

training have verified this fact. Therefore,

these proposed changes do not alter or affect

any UFSAR or off-site dose calculation

assumptions, and the margin of safety is not

reduced.

A review of the Byron and Braidwood PRA

shows that these proposed changes will have

no effect on either CDF or URF.

No new equipment is being installed, and

no existing equipment is being modified. The

VA system will continue to operate as

designed during normal and post accident

conditions. All of the accident analysis

assumptions remain satisfied.

Therefore this proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O.

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603.

NRC Project Director: Robert A. Capra.
Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices 11489

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

455, Byron Station, Unit Nos. 1 and 2,

Ogle County, Illinois

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1 and

2, Will County, Illinois

Date of amendment request: January

20, 1997.

Description of amendment request:

The proposed amendment would

change Technical Specification Table

3.6–1 to reflect planned changes in the

plant configuration. As a result of the

planned replacement of the

Westinghouse D4 steam generators at

Byron, Unit 1, and Braidwood, Unit 1,

changes will be made to the

containment isolation piping

arrangements at the penetrations

associated with the Feedwater (FW) and

Auxiliary Feedwater (AF) systems. As a

result of these changes, there will be no

split FW flow with the replacement

steam generators. AF flow will be fed

into the main FW piping outside of

containment and the existing FW

tempering penetration will be used for

a new steam generator recirculation

system to be used during periods of

extended shutdown. Additionally, since

the replacement steam generators use a

feedring design rather than a preheater

design, the FW Isolation Bypass line

and associated containment isolation

valves will no longer be required. Table

3.6–1 of the Technical Specifications

(TS) must be updated to reflect these

changes. These changes do not affect the

containment isolation capability

originally designed to the criteria in 10

CFR 50, Appendix A, General Design

Criteria (GDC) 54 through 57 as reflected

in the Byron/Braidwood Updated Final

Safety Analysis Report (UFSAR).

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

Technical Specification 3/4.6.3 establishes

the operability requirements for containment

isolation valves as required by the Byron and

Braidwood Operating Licenses in compliance

with General Design Criteria 54 through 57

of Appendix A to 10 CFR 50. The operability

of the containment isolation valves ensure

that the containment atmosphere will be

isolated from the outside environment in the

event of a release of radioactive material to

the containment atmosphere. Table 3.6–1

identifies these isolation valves and captures

relevant information to ensure these valves

remain operable under required conditions.

These proposed changes result in the

elimination of the FW Isolation Bypass

isolation valves. These isolation valves are

not required with the replacement steam

generator design. The remaining isolation

valves have not been altered in any way, only

the piping associated with them has been

altered to the revised configuration. These

changes do not result in alteration of any

containment penetrations.

Failure of the piping between the isolation

valve and the containment penetration is

considered as an accident initiator. However,

all piping changes between the isolation

valve and the containment penetrations meet

the requirements of the original design.

Therefore, since all original piping design

criteria are met and the actual number of

containment isolation valves is reduced, the

proposed change does not involve a

significant increase in the probability of an

accident previously evaluated.

Each penetration identified in the

proposed change is associated with a closed

system inside containment and, as such, is

provided containment isolation in

accordance with the applicable requirements

of GDC 54 through 57. There are four

analyzed transients which take credit for

feedwater isolation and are, therefore,

relevant to this proposed change. These

accidents are: (1) feedwater system

malfunctions that result in an increase in FW

flow, (2) inadvertent opening of a steam

generator relief or safety valve, (3) steam

system piping failure, and (4) FW system

pipe break. All operability requirements for

the affected containment isolation valves are

unaffected by this proposed change.

The containment isolation valves’

functions, system operating conditions, and

accident responses are unchanged as a result

of the new configuration. Therefore, since all

original design criteria are met and each

remaining isolation valve continues to

provide the same degree of containment

isolation as the original design, the proposed

change does 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.

All modifications associated with the

proposed changes will be outside of

containment and can be characterized as the

rearrangement of piping systems. All piping

changes will comply with the original design

of the plant and will retain required

containment isolation capabilities per the

requirements of GDC 54 through 57 as

required by the current design basis. Piping

configurations within the area of the

containment penetration and the

containment isolation valves are required to

minimize branch connections per guidance

in the Standard Review Plan (SRP) Section

3.6.2.

Therefore, since there are no unique

configurations or reductions in design

requirements, this proposed change does not

create the possibility of any new or different

kinds of accidents from those previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed changes to the containment

isolation arrangement are being made

consistent with the same codes, standards,

and isolation criteria as are currently in use

at Byron and Braidwood. The containment

isolation valves remaining in place following

the steam generator replacement are

unchanged with regard to their function,

capability, reliability, or physical

requirements. Containment isolation

capability in accordance with GDC 54

through 57 is maintained at current levels of

protection for the health and safety of the

general public. Therefore, this proposed

change does not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O.

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

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. STN 50–454 and STN 50–

455, Byron Station, Unit Nos. 1 and 2,

Ogle County, Illinois

Date of amendment request: January

31, 1997.

Description of amendment request:

The proposed amendments would

revise the maximum allowable value in

the Byron, Unit 1, Technical

Specifications (TS), of the dose

equivalent (DE) iodine-131

concentration in the primary coolant

from the present value of 0.35

microcuries per gram of coolant to a

maximum allowable of 0.20 microcuries

per gram. This reduction in the DE

iodine-131 concentration would be

applicable only for the remainder of the

present Byron, Unit 1, operating cycle

(i.e., fuel cycle 8) which the licensee has

previously stated will end in December

1997. The subject amendments are

proposed by the licensee in order to

provide additional margin with respect

to the maximum Byron Station site

allowable primary-to-secondary leakage

limit from the Byron, Unit 1, steam

generators (SG). This proposed Byron,

Unit 1, TS revision to increase this

margin is being proposed in conjunction
11490 Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices

with the proposed operating interval of

540 days above a Thot temperature of 500

degrees Fahrenheit, between eddy

current inspections (ECI) of the Byron 1

SGs. The last Byron, Unit 1, ECI was

initiated in November 1995. This

margin increase is being sought by the

licensee to address staff concerns

regarding potential SG tube leakage

under postulated accident conditions

due to SG tube circumferential cracking

at the top of the tubesheet in the roll

transition zone.

While the proposed revision to the DE

iodine-131 is applicable only to Byron,

Unit 1, the pending request for license

amendments involves both Byron, Units

1 and 2, in that both units have a

common set of TSs.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

Generic Letter 95–05, ‘‘Voltage Based

Repair Criteria For Westinghouse Steam

Generator Tubes Affected By Outside

Diameter Stress Corrosion Cracking,’’ allows

lowering of the RCS DE I–131 activity as a

means for accepting higher projected leak

rates if justification for equivalent I–131

below 0.35 microcuries/gm is provided. Four

methods for determining the impact of a

release of activity to the public were

reviewed to provide the justification.

They are as follows:

Method 1: NRC NUREG 0800, Standard

Review Plan (SRP) Methodology

Method 2: Methodology described in a report

by J.P. Adams and C.L. Atwood, ‘‘The

Iodine Spike Release Rate During a Steam

Generator Tube Rupture,’’ Nuclear

Technology, Vol. 94 p. 361 (1991), using

Byron Station reactor trip data.

Method 3: Methodology described in Adams

and Atwood report, using normalized

industry reactor trip data.

Method 4: Methodology described in draft

EPRI Report TR–103680, Revision 1,

November 1995, ‘‘Empirical Study of

Iodine Spiking in PWR Plants’’.

The effect of reducing the RCS DE I–131

limit on the amount of activity released to the

environment remains unchanged when the

maximum site allowable primary-tosecondary leakage limit is proportionately

increased. With a DE I–131 limit of 1.0

microcuries/gm, the maximum site allowable

leakage limit was calculated in accordance

with the NRC SRP methodology to be 12.8

gpm. The corresponding calculated activity

released during a MSLB is 15.8 Ci. ComEd

has evaluated the reduction of the DE I–131

to 0.20 microcuries/gm along with the

increase of the allowable leakage to 64 gpm

and has concluded:

—The maximum activity released is not

changed, and

—The offsite dose including the iodine

spiking factor is bounded by method 1.

Therefore, the offsite dose assessment and

conclusions previously reached remain valid

and continue to meet the requirements of 10

CFR 100.

An evaluation of Control Room dose

attributed to a MSLB concurrent with steam

generator primary-to-secondary leakage at the

site allowable leakage limit was performed in

support of a license amendment request for

application of 1.0 volt Interim Plugging

Criteria. This evaluation concluded that

Control Room dose due to the MSLB scenario

is bounded by the existing loss of coolant

accident analysis. Therefore, the maximum

site allowable primary-to-secondary leakage

limit continues to be based on offsite dose at

the Exclusion Area boundary due to MSLB

leakage. This conclusion was previously

submitted to the Staff in a September 22,

1994, transmittal in support of the 1.0 volt

Interim Plugging Criteria license amendment

request.

Based on the NRC SRP methodology for

dose assessments, the Control Room dose, the

Low Population Zone dose, and the dose at

the Exclusion Area Boundary continue to

satisfy the appropriate fraction of the

10CFR100 dose limits.

The Adams and Atwood report concluded

that the NRC SRP methodology, which

specifies a release rate spike factor of 500 for

iodine activity from the fuel rod to the RCS,

is conservative. In order to justify that a

release rate spike factor of 500 is

conservative, actual operating data from the

previous reactor trips of Byron Unit 1 and

Unit 2, with and without fuel failures, were

reviewed and analyzed using the

methodology presented Section II.C of the

Adams and Atwood report (Method 2). The

same five data screening criteria described in

the Adams and Atwood report were applied

to the Byron data to ensure consistency and

validity when comparing the Byron results to

the data in the Adams and Atwood report. Of

the twenty-eight (28) reactor trip events at

Byron Units 1 and 2, twelve (12) met the five

data screening criteria.

Three of the Byron trips occurred during

cycles with no failed fuel. In all three of these

instances, the calculated spike factor was less

than the spike factor of 500 assumed in the

NRC SRP methodology. Byron, Unit 1, Cycle

8 is currently operating with no failed fuel

and a DE I–131 activity of approximately 6E–

4 microcuries/gm. The three previous trips

with no fuel failures had steady-state iodine

values that are relatively close to current

operating conditions. It is therefore

reasonable to conclude that the calculated

spike factors from those trips would reflect

the spike factor expected from an actual trip

during the current cycle.

Based on the data in the Adams and

Atwood report, the NRC SRP release rate

spike factor of 500 may seem nonconservative since the Adams and Atwood

factor was typically greater than 500 when

initial concentrations were less than 0.3

microcuries/gm. The primary reason for these

high ratios (up to 12,000) is not because the

absolute post-trip release rate is high (factor

numerator), but rather because the steadystate release rate (factor denominator) is low.

The Byron specific data only resulted in one

trip with a calculated release rate spike factor

greater than 500, a value of 603.9. The trip

occurred during the first operating cycle of

Unit 2 which experienced failed fuel and a

very low steady-state release rate. It is not

expected based upon the current fuel cycle

conditions that a spiking factor of greater

than 500 would occur.

In order to compare the Byron specific data

to the NRC SRP methodology, the release rate

for a steady-state RCS DE I–131 activity of 1.0

microcuries/gm was calculated. Using the

Byron specific data, the steady-state release

rate is 17.6 Ci/hr. Using a release rate factor

of 500 for the accident initiated spike, the

post-trip maximum release rate would be

8797 Ci/hr. This is significantly higher than

the largest iodine release rate of 127 Ci/hr

from the Byron data. This demonstrates that,

although a data point shows an iodine spike

factor greater than 500, the resulting post-trip

RCS DE I–131 fuel rod iodine release rate is

less than the fuel rod iodine release rate from

the NRC SRP methodology.

In the fourth method, the results from Draft

EPRI Report TR–103680, Rev. 1, November

1995, ‘‘Empirical Study of Iodine Spiking In

PWR Power Plants’’ were applied. The

objective of the EPRI study was to quantify

the iodine spiking in postulated Main Steam

Line Break/Steam Generator Tube Rupture

(MSLB/SGTR) sequences. In the EPRI report,

an iodine spike factor between 40 and 150

was determined to match data from existing

plant trips. The maximum iodine spike factor

value of 150 was applied to a steady-state

equilibrium RCS DE I–131 activity of 0.33

microcuries/gm. The resulting 2-hour average

iodine concentration for a postulated MSLB/

SGTR sequence was determined to be 3.1

microcuries/gm. Since the EPRI report is

based on industry data and the EPRI method

predicted a post-accident iodine activity

which is a small fraction of the activity

predicted by the NRC SRP methodology, it

can be expected that, for the proposed 0.2

microcuries/gm limit under a MSLB/SGTR

sequence, the post-accident iodine activity

would be a small fraction of the RCS DE I–

131 activity predicted by the NRC SRP

methodology.

Lowering the Unit 1 RCS DE I–131 activity

limit is conservative and remains bounded by

the NRC SRP methodology. Thus, all offsite

and control room dose assessment

conclusions satisfy the appropriate limits of

10 CFR 100 and GDC 19. These proposed

changes do not result in a significant increase

in the consequences of an accident

previously analyzed.

The RCS DE I–131 activity limit is not

considered as a precursor to any accident.

Therefore, this proposed change does not

result in a significant increase in the

probability of an accident previously

analyzed.

The correction of the typographical error is

administrative in nature and has no impact

on either the probability or consequences of

an accident previously analyzed.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.
Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices 11491

The changes proposed in this amendment

request conservatively reduce the Unit 1 DE

I–131 limit at which action needs to be taken

and correct a typographical error. The

changes do not directly affect plant

operation. These changes will not result in

the installation of any new equipment or

systems or the modification of any existing

equipment or systems. No new operating

procedures, conditions or modes will be

created by this proposed amendment.

Thus, this proposed change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

NRC Generic Letter 95–05 allows lowering

of the dose equivalent iodine as a means for

accepting higher projected leakage rates

provided justification for equivalent I–131

below 0.35 microcuries/gm is provided. Four

methods for determining the fuel rod iodine

release rates and spike factors during an

accident were reviewed. Each of these

methods utilized actual industry data,

including Byron, Unit 1 and Unit 2, for preand post-reactor trip DE I–131 activities.

Each of the methods demonstrated that the

actual fuel rod iodine release rates are a small

fraction of the release rate as calculated using

the NRC SRP methodology. All design basis

and off-site dose calculation assumptions

remain satisfied. This proposed change will

not result in a reduction in a margin of safety.

Correction of the typographical error is

administrative in nature and does not impact

the margin of safety. Therefore, the proposed

changes do not result in a significant

reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: Byron Public Library District,

109 N. Franklin, P.O. Box 434, Byron,

Illinois 61010.

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. STN 50–454 and STN 50–

455, Byron Station, Unit Nos. 1 and 2,

Ogle County, Illinois

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1 and

2, Will County, Illinois

Date of amendment request: February

18, 1997.

Description of amendment request:

The proposed amendment would revise

Byron and Braidwood Technical

Specification (TS) Table 2.2–1

(functional unit 13.a), ‘‘Reactor Trip

System Instrumentation Trip Setpoint:

Steam Generator Water Level Low-Low’;

TS Table 3.3–4 (functional unit 5.b.1),

‘‘Engineered Safety Features Actuation

System Instrumentation Trip Setpoints:

Steam Generator Water Level-HighHigh’; TS Table 3.3–4 (6.c.1),

‘‘Engineered Safety Features Actuation

System Instrumentation Trip Setpoints:

Steam Generator Water Level-Low-Low

Start Motor-Driven Pump and DieselDriven Pump’; TS Surveillance

Requirement (TSSR) 4.4.1.2.2, required

steam generator inventory during hot

standby; TSSR 4.4.1.3.2, required steam

generator inventory during hot

shutdown; and TS Section 3.4.1.4.1.b,

limiting condition for operation during

cold shutdown with loops filled.

The installation of Babcock and

Wilcox International (BWI), replacement

steam generators (RSGs) at Byron, Unit

1, and Braidwood, Unit 1, necessitates

an increase to the operating range of the

steam generators due to the decrease in

narrow range span from 233 inches for

the original Westinghouse Model D4

steam generators (OSGs) to 180 inches

for the BWI RSGs. The increase in

operating range will minimize the

possibility of inadvertent plant trips

following load changes and feedwater

transients.

ComEd also proposes to eliminate

notations from page 2–5 for both

Braidwood and Byron and pages 3/4 3–

25 and 3/4 3–26 (for Braidwood only)

since they are related to cycles already

completed and, therefore, are no longer

valid.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

This proposed change includes changing

the low-low and high-high SG level

setpoints. The setpoints are being changed to

increase the SG level operating range. The

change in acceptable operating range will

decrease the possibility of inadvertent plant

trips following load changes and feedwater

transients. Therefore, the probability of

inadvertent plant trips will decrease with

this change.

The minimum setpoint change proposed in

this request establishes controls to ensure

that an adequate heat sink is maintained by

providing an adequate secondary liquid mass

to remove primary system sensible heat and

core decay heat shortly after reactor trip and

initiating auxiliary feedwater flow for longterm cooling. The accidents evaluated for this

requirement are the Loss of Normal

Feedwater and Feedwater Line Break

transients.

The maximum setpoint ensures the steam

lines and turbine remain undamaged from

the introduction of low quality, two-phase

flow from the steam generators into the steam

lines. The accident evaluated for this

requirement is the Feedwater System

Malfunction that results in an increase in

feedwater to one or more steam generators.

The steam generator water level setpoints

are not considered a precursor to any of the

analayzed accidents, and, therefore, these

proposed changes do not result in an increase

in the probability of occurrence of any

accident previously analyzed.

The accidents evaluated for the low-low

setpoint are the Loss of Normal Feedwater

and Feedwater Line Break transients. These

accidents were both analyzed using approved

methodologies. All acceptance criteria were

shown to be met for both these events. In

addition, it was demonstrated that the

Feedwater System Pipe Break response with

the RSGs and the proposed low-low setpoint

were bounded by the response with the

original Model D4 steam generators.

Therefore, the proposed low-low level

setpoint change is demonstrated not to result

in an increase in the consequences for these

accidents.

The accident evaluated for the high-high

setpoint is the Feedwater System

Malfunction that results in an increase in

feedwater to one or more Steam Generators.

All acceptance criteria were shown to be met.

In addition, it was shown that the RSGs do

not completely fill with liquid. This assures

that the steam lines and turbine remain

undamaged with no introduction of low

quality, two-phase flow from the steam

generators into the steam lines during the

transient. With all acceptance criteria met,

the proposed high-high level setpoint change

is demonstrated not to result in an increase

in the consequences for these accidents.

TSSR 4.4.1.2.2, TSSR 4.4.1.3.2, and TS

3.4.1.4.1.b assure a minimum inventory (i.e.,

level) to provide decay heat removal. The

requirement for a minimum inventory to

remove decay heat is met with assurance that

the tube bundle is completely covered. The

steam generator operating water level during

shutdown conditions are not considered a

precursor to any accident, and, therefore,

these proposed changes do not result in an

increase in the probability of occurrence of

any accident previously analyzed.

The elimination of outdated cycle specific

notations from page 2–5 for both Braidwood

and Byron and pages 3/4 3–25 and 3/4 3–26

(Braidwood only) are only administrative and

does not impact the probability or

consequences of any accidents previously

analyzed.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed setpoint changes do not

create any new operating conditions or

modes. The proposed change only revises the

setpoints for the Reactor Trip System and

Engineered Safety Features Actuation

System. The actions of these systems will

continue to be performed in accordance with
11492 Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices

existing requirements which are sufficient to

ensure plant safety is maintained.

Shutdown conditions steam generator

water level is necessary to assure adequate

decay heat removal capacity. Assurance that

the tube bundle is completely covered along

with existing technical specification controls

on the Auxiliary Feedwater System and on

the Condensate Storage Tank ensure

adequate heat removal capacity is maintained

and that plant safety is maintained.

Thus, this proposed change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The elimination of outdated cycle specific

notations from page 2–5 for both Braidwood

and Byron and pages 3/4 3–25 and 3/4 3–26

(Braidwood only) are only administrative and

does not create the possibility of a new or

different accident.

3. The proposed change does not involve

a significant reduction in a margin of safety.

A safety evaluation was performed to

determine the effect of the RSGs with the

revised setpoints.

The accidents potentially affected by the

change in the Reactor Trip Steam Generator

Water Level low-low setpoint (TS 2.2.1,

Table 2.2–1, functional unit 13.a) and

Engineered Safety Features Actuation System

low-low AFW start setpoint (TS 3.3.2, Table

3.3–4, functional unit 6.c.1) are the Loss of

Normal Feedwater and Feedwater Line Break

transients. These accidents were both

analyzed using approved methodologies. All

acceptance criteria were shown to be met for

both these events.

In addition, it was demonstrated that the

Feedwater System Pipe Break response with

the RSGs with the proposed low-low setpoint

were bounded by the response with the

OSGs. Therefore, the proposed low-low level

setpoint change is demonstrated not to result

in an reduction in the margin of safety for

these accidents.

The accident potentially affected by the

change in the Engineered Safety Features

Actuation System high-high SG level trip (TS

3.3.2, Table 3.3–4, functional unit 5.b.1) is a

Feedwater System Malfunction that results in

an increase in feedwater to one or more

steam generators. This accident was analyzed

using an approved methodology. In the

evaluation of the Feedwater System

Malfunction, all acceptance criteria were

shown to be met. In addition, it was shown

that the RSGs do not completely fill with

liquid. This assures that the steam lines and

turbine remain undamaged with no

introduction of low quality, two-phase flow

from the steam generators into the steam

lines during the transient. With all

acceptance criteria met, the proposed highhigh level setpoint change is demonstrated

not to result in a reduction in the margin of

safety.

There are no design basis accidents

involving shutdown condition steam

generator water level. Existing TS controls on

the Auxiliary Feedwater System and on the

Condensate Storage Tank ensure adequate

heat removal capacity is maintained and that

plant safety is maintained during shutdown

conditions. Therefore, a change to the

shutdown condition steam generator water

level does not result in a reduction in the

margin of safety.

The elimination of outdated cycle specific

notations from page 2–5 for both Braidwood

and Byron and pages 3/4 3–25 and 3/4 3–26

(for Braidwood only) are only administrative

and does not result in a reduction in the

margin of safety for any analyzed event.

Therefore, this amendment request does

not result in a significant decrease 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

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O.

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603

NRC Project Director: Robert A. Capra.

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

No. 1, Lake County, Ohio

Date of amendment request: January

31, 1997.

Description of amendment request:

The proposed amendment will insert,

by general reference, in the Perry

Nuclear Power Plant Technical

Specifications, the implementation

document that the licensee will use to

implement Option B, ‘‘PerformanceBased Requirements,’’ to 10 CFR 50,

Appendix J, ‘‘Primary Reactor

Containment Leakage Testing for WaterCooled Power Reactors.’’ Option B to 10

CFR 50 Appendix J is an option that

became effective on October 26, 1995.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration which is presented below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The changes involved in this license

amendment request revise the criteria for

determining the Containment leak rate

testing interval based upon past component

performance. The revised criteria are based

on the guidance contained in Regulatory

Guide 1.163, ‘‘Performance-Based

Containment Leak-Test Program.’’ When the

containment or containment penetrations

have performed satisfactorily on a historical

basis, this guidance permits the use of

extended testing frequencies.

Since the allowable leakage rates are not

being affected, the performance of the

primary containment and systems and

components penetrating the primary

containment remains within acceptable

limits. The functions and operation of these

components will remain unchanged. Since

the components are utilized to mitigate the

consequences of accidents that require

containment isolation, they are not

considered to be accident initiators.

Additionally, there are no accidents

associated with implementation of a

performance-based testing frequency for the

primary containment and systems and

components penetrating the primary

containment.

As discussed previously, the components

are utilized to mitigate the consequences of

accident scenarios which rely upon the

primary containment and systems and

components penetrating the primary

containment, to prevent the release of

radioactive effluents. The implementation of

Option B to 10 CFR 50 Appendix J is not

intended to provide relief from the leakage

criteria. The components will still be

required to meet the leakage requirements as

discussed in USAR Section 6.2.6 and

Technical Specifications 3.6.1.1, 3.6.1.2, and

3.6.1.3. The primary containment isolation

system is designed to limit leakage to La,

which is defined by the Perry Technical

Specifications to be 0.20 percent of primary

containment air weight per day at the

calculated peak containment pressure (Pa) for

the design basis loss of coolant accident. The

limitation on the rate of primary containment

leakage is designed to ensure that the total

leakage volume will not exceed the value

assumed in the accident analyses at Pa. The

La value is not being modified by this

proposed change. Based on this, the primary

containment and system and components

penetrating the primary containment will

remain capable of maintaining radioactive

effluent releases within the limits of 10 CFR

100.

Because the proposed change does not alter

the plant design, including the primary

containment and primary containment

penetrations, the proposed change does not

directly result in an increase in primary

containment leakage. Since the frequency

will be based on the performance of the

subject components, only those components

that have satisfactorily maintained the actual

leakage less than the allowable leakage will

be tested less frequently. The testing

frequency for components which have not

satisfactorily limited leakage, or have not

performed satisfactorily in the past, will not

be altered. Other programs are also in place

to ensure that proper maintenance and

repairs are performed during the service life

of the primary containment and systems and

components penetrating the primary

containment.
Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices 11493

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of previously

evaluated accidents.

Several administrative/editorial changes

have been incorporated (e.g., the clarification

of the ‘‘less than’’ and ‘‘less than or equal to’’

signs on the Technical Specification

acceptance criteria, and the retention of the

standard frequency for the Drywell visual

inspections). Such administrative/editorial

changes do not impact initiators of analyzed

events or assumed mitigation of accident or

transient events. Therefore, these changes

also do not involve a significant increase in

the probability or consequences of an

accident previously evaluated.

2. The proposed change would not create

the possibility of a new or different kind of

accident from any previously evaluated.

The proposed change does not involve a

change to the plant design or operation, or

new system interfaces. Consequently, the

proposed change does not affect the

parameters or conditions that could

contribute to initiation of accidents. This

change involves adopting a performancebased method for determining Type A, B, and

C test frequencies. Except for the method of

defining the test frequency, the methods for

performing the actual tests are not changed.

No new accident modes would be created by

extending testing intervals. No safety related

equipment or safety functions are altered as

a result of this change. The change in testing

frequency will not create any different types

of accidents since the primary containment

and systems and components penetrating the

primary containment will continue to operate

within their design bases. Therefore,

reducing the test frequency would have no

influence on, nor contribute to, the

possibility of a new or different kind of

accident or malfunction from those

previously analyzed.

Based on the above discussions, the

proposed change would not create the

possibility of a new or different kind of

accident than those previously evaluated.

The proposed administrative/editorial

changes do not involve a physical alteration

of the plant (no new or different type of

equipment will be installed) or changes in

methods governing normal plant operation.

Thus, these changes also do not create the

possibility of a new or different kind of

accident from any previously evaluated.

3. The proposed change will not involve a

significant reduction in the margin of safety.

This request does not involve a significant

reduction in a margin of safety. The proposed

change adopts a performance-based method

for determining frequency of Type A, B, and

C testing.

Except for the method of defining test

frequency, no change in the method of testing

is proposed. Since the frequency will be

based on the performance of the subject

components, only those components that

have satisfactorily maintained actual leakage

less than the allowable leakage will be tested

less frequently. Other programs are also in

place to ensure that proper maintenance and

repairs are performed during the service life

of the primary containment and systems and

components penetrating the primary

containment.

The margin of safety associated with the

proposed change involves the offsite dose

consequences of postulated accidents, which

are directly related to the rate of primary

containment leakage. The primary

containment isolation system is designed to

limit leakage to La, which is defined by the

Perry Technical Specifications to be 0.20

percent of primary containment air weight

per day at the calculated peak containment

pressure (Pa) for the design basis loss of

coolant accident. The limitation on the rate

of primary containment leakage is designed

to ensure that the total leakage volume will

not exceed the value assumed in the accident

analyses at Pa. The margin of safety for the

offsite dose consequences of postulated

accidents directly related to the primary

containment leakage rate is maintained by

continuing to meet La. The La value is not

being modified by this proposed change.

Based on this, the primary containment and

systems and components penetrating the

primary containment will remain capable of

maintaining radioactive effluent releases

within the limits of 10 CFR 100.

Therefore, the changes associated with this

license amendment request do not involve a

significant reduction in the margin of safety.

The proposed administrative/editorial

changes will not reduce the margin of safety

because they have no impact on safety

analysis assumptions. These changes do not

involve questions regarding safety issues, and

therefore also 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: Perry Public Library, 3753

Main Street, Perry, Ohio 44081.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts & Trowbridge,

2300 N Street, NW., Washington, DC

20037.

NRC Project Director: Gail H. Marcus.

Dairyland Power Cooperative (DPC),

Docket No. 50–409, LaCrosse Boiling

Water Reactor (LACBWR), Vernon

County, Wisconsin

Date of amendment request: April 10,

1996.

Description of amendment request:

This is a corrected notice that was first

issued on August 1, 1996. The proposed

amendment would update the facility

Possession Only License and Technical

Specifications to reflect the permanently

shutdown and defueled condition of the

plant. The amendment would also serve

to remove the fire protection

requirements, radiological effluent

controls, quality assurance program

controls and administrative controls for

the emergency and security plans from

the Technical Specifications to other

inspectable and enforceable documents.

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:

DPC proposes to modify the LACBWR

Technical Specifications to more accurately

reflect the permanently shutdown, defueled,

possession-only status of the facility.

Analysis of no significant hazards

consideration:

1. The proposed changes do not create a

significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes delete system

requirements that are no longer necessary to

prevent, or mitigate the consequences of, a

credible SAFSTOR accident as described in

our current SAFSTOR Accident Analysis.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed changes are either

administrative in nature or were made based

on the analysis of previously evaluated

accident scenarios. In no other way do they

change the design or operation of the facility

and therefore do not create the possibility of

a new or different kind of accident from any

previously evaluated.

3. The proposed changes do not result in

a significant reduction in the margin of

safety.

The changes incorporate into the proposed

Technical Specifications the margin of safety

associated with the current SAFSTOR

accident analysis and thus don’t involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis, and based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: LaCrosse Public Library, 800

Main Street, LaCrosse, Wisconsin

54601.

Attorney for licensee: Wheeler, Van

Sickle and Anderson, Suite 801, 25

West Main Street, Madison, Wisconsin

53703–3398.

NRC Project Director: Seymour H.

Weiss.

Duke Power Company, 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: February

10, 1997 (TSC 95–04).

Description of amendment request:

The proposed changes would revise the
11494 Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices

Technical Specifications (TS) to reduce

the allowable reactor building volume

leakage rate per-day limit to permit

removal of consideration of the

penetration room contribution to the

limit and the requirement to maintain

the penetration room at a negative

pressure with respect to all adjacent

areas. Also, the penetration room

ventilation system would be removed

from the description of the containment

in TS 5.2, and a surveillance

requirement to perform a refueling

outage test of the penetration room

ventilation system would be added to

TS 4.5.4. In addition, related changes

would be made to the appropriate Bases

sections.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

A. Involve a significant increase in the

probability or consequences of an accident

previously evaluated?

No.

The following requirements are being

removed from Technical Specifications

regarding the PRVS [Penetration Room

Ventilation System]:

(1) The requirement to measure reactor

building leakage in excess of 50% of the total

allowed containment leakage to the

penetration room.

(2) The requirement, as specified in the

design features, for the PRVS to maintain the

penetration room at a negative pressure with

respect to all adjacent areas. In addition, the

design features description for the PRVS will

be completely removed from Technical

Specification 5.2 and replaced with a

surveillance requirement in Technical

Specification 4.5.4.

To demonstrate the inconsequential effects

of the removal of the above requirements, a

dose analysis was performed to

conservatively demonstrate that PRVS adds

margin, but is not necessary to meet

10CFR100 limits. The analysis assumes that

the PRVS is completely unavailable for

offsite dose reduction. However, the PRVS

will be available, and all of the relevant

operability and surveillance requirements for

the PRVS will be retained in the Technical

Specifications. Therefore, it is highly

unlikely that the actual dose consequences

would increase from 167 Rem thyroid to 240

Rem thyroid, since all surveillance and

operability requirements for PRVS, other

than the two requirements specified above,

will be retained in Technical Specifications.

The specified Technical Specification

requirements for PRVS are not accident

initiators, nor will these requirements impact

the probability of an accident. The purpose

of these requirements is to ensure that the

PRVS can reduce offsite dose to the public

in the event of an accident which results in

radioactive effluents leaking from the Reactor

Building (RB) into the Penetration Room

(PR).

In the initial ONS [Oconee Nuclear Station]

design basis, the PRVS was credited to

reduce offsite dose to the public in the event

of certain accidents, such as a loss of coolant

accident (LOCA) or Maximum Hypothetical

Accident (MHA), where there is airborne

leakage of radioactivity from the RB into the

PR. The PRVS was credited to reduce the

MHA two-hour Exclusion Area Boundary

(EAB) dose to less than the 10CFR100 limit

of 300 Rem thyroid. The current ONS dose

analysis, which takes credit for the PRVS,

calculates the MHA two-hour EAB dose to be

167 Rem thyroid. With a reduction in the

allowable leakage from the Reactor Building

(La) from 0.25 w%/day to 0.20 w%/day,

while taking no credit for the PRVS, the two

hour EAB MHA dose is calculated to be 240

Rem thyroid. This new dose analysis result

meets the acceptance criterion of 10CFR100.

In addition to conducting a detailed dose

analysis without taking credit for PRVS, a

detailed review of PRA [probabilistic risk

analysis] risk significance of the PRVS was

conducted. The PRVS was determined to

have virtually no PRA risk significance and

no significant impact on consequences.

A review of the impact on control room

habitability due to the proposed Technical

Specification changes was conducted for

credible UFSAR [Updated Final Safety

Analysis Report] Chapter 15 accident

scenarios. The operability requirements of

the PRVS which are being retained in the

Technical Specifications will ensure

operability requirements are met to support

the Control Room Ventilation System

(CRVS). Therefore, removal of the identified

statements pertaining to PRVS operability

from Technical Specifications will not

significantly impact control room

habitability.

Based on the above information, the

removal of the specified requirements for

PRVS from Technical Specifications will not

significantly increase the probability or

consequences of an accident previously

evaluated. The original design basis for

offsite dose will still be met without any

credit taken for the PRVS.

A change has been proposed to the

Technical Specifications to reduce the

allowable leakage from the Reactor Building

(La) from 0.25 w%/day to 0.20 w%/day. This

proposed change is conservative in nature

since it will result in a potential reduction in

the consequences of any accidents previously

evaluated. Past integrated leak rate tests

(ILRTs) for all three Oconee units have been

reviewed by engineering and it has been

concluded that this reduction in allowable

leakage will have no impact on future station

operation. This reduction is possible since

the actual leakage of the ONS reactor

buildings is far less than the original

allowable design leakage.

B. Create the possibility of a new or

different kind of accident from the accident

previously evaluated?

No.

As stated previously, the proposed

Technical Specification changes for the PRVS

are not accident initiators, nor will these

changes create the possibility of new or

different kinds of accidents. The purpose of

the PRVS is to reduce offsite dose to the

public in the event of an accident which

results in leakage from the RB into the PR.

Therefore, the proposed changes to the

Technical Specifications will not create the

possibility of a new or different kind of

accident from the accidents previously

evaluated.

C. Involve a significant reduction in a

margin of safety?

No.

By reducing the allowable La to 0.20 w%/

day, ONS meets 10CFR100 limits for off-site

dose without taking any credit for the PRVS.

Although the margin to 10CFR100 limits is

reduced by not taking credit for PRVS, it is

concluded that the reduction in margin of

safety is insignificant because:

(1) PRVS operability and surveillance

requirements are being retained in Technical

Specifications with the exception of two

items which do not significantly degrade the

ability of PRVS to perform its function.

(2) The reduction in the margin of safety

is being offset by a reduction in La.

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: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691.

GPU Nuclear, Inc. and Saxton Nuclear

Experimental Corporation, Docket No.

50–146, Saxton Nuclear Experimental

Facility (SNEF), Bedford County,

Pennsylvania

Date of amendment request:

November 25, 1996.

Description of amendment request:

The proposed amendment would allow

decommissioning of the SNEF. The

proposed changes to the license and

technical specifications (TSs) would (1)

accommodate decommissioning

activities at the SNEF, (2) establish

specific TS controls such as

administrative controls and inspection

requirements over decommissioning

activities, (3) establish limiting

conditions for performing

decommissioning activities, (4) extend

exclusion area controls to include the

SNEF Decommissioning Support

Building, (5) establish requirements for

a Radiological Environmental

Monitoring Program, an Off-Site Dose

Calculation Manual and a Process

Control Program, and (6) establish

requirements for Technical and

Independent Safety Reviews. In

addition, the licensees have proposed

other administrative and editorial
Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices 11495

changes to the TSs associated with the

changes proposed above.

Basis for Proposed No Significant

Hazards Consideration Determination:

As required by 10 CFR 50.91(a), the

licensees have provided their analysis of

the issue of no significant hazards

consideration, which is presented

below:

The proposed changes do not involve a

significant hazards consideration because the

changes would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Accidents which might occur during the

active decommissioning phase of the SNEF

are bounded by the twelve accidents

addressed in section 3.0 of the Updated

Safety Analysis Report (USAR). The accident

analyses addressed in the USAR demonstrate

that no adverse public health and safety

impacts are expected from accidents that

might occur during decommissioning

operations at the SNEF. The highest

calculated dose to an individual located at

the site boundary is less than 1.5 mrem to the

whole body during a postulated materials

handling accident. The dose to an individual

located at the site boundary for other on-site

accidents is at or below this value. The

limiting accident case represents less than

0.15% of the EPA lower whole body dose

limit for radiological accidents. Based on the

analyses of postulated credible accidents that

might occur during the planned

decommissioning operations at the SNEF, it

is concluded that no significant increase in

the probability or consequences of an

accident previously evaluated would be

involved.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

There are three general categories of

accidents. These scenarios evaluate different

methods of dispersing radioactive material to

the environment which include a loss of

support systems and external events. The

first includes accident scenarios associated

with decommissioning tasks. These were

identified and evaluated as described in

Section 3.0 of the USAR. The radiological

effects of these accident scenarios are

discussed in item 1 above. They do not,

therefore, reflect a new or different kind of

accident previously evaluated. The second

category, loss of support systems, does not

directly lead to an accident situation.

Therefore, this category of event does not

create the possibility of a new or different

kind of accident. The final category of

accidents involves external events.

Since these types of events can occur

whether the SNEF is being decommissioned

or not, the act of decommissioning does not

create the possibility of a new or different

kind of external event. Any potential

radiological hazard that may occur as a result

of an external event is addressed in item 1

above.

3. Involve a significant reduction in a

margin of safety.

The TSs currently in place at the SNEF

were developed to maintain a shutdown

facility in a secured condition with

occasional monitoring. These specifications

were designed to ensure that the

approximately 4 megacuries of radioactive

material left on site following shutdown in

1972 as identified in the Saxton

Decommissioning Plan and Safety Analysis

Report dated April 1972, would remain

safely contained. In the ensuing years,

natural decay of these radioactive materials

has resulted in a remainder of approximately

1500 curies of radioactive material at the

facility (93% of which is activation contained

within the steel structures of the reactor

vessel). These proposed decommissioning

TSs were developed in order to ensure this

remaining radioactive material is safely

contained and disposed of and that the

environment surrounding the facility is

monitored. These actions will assure that

there is no reduction in the margin of safety

during the active decommissioning of the

facility. The final result of these efforts will

be the removal of any potential radiological

hazard from the site and the release of the

site for unrestricted use.

The NRC staff has reviewed the

analysis of the licensees and, based on

this review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

Location: Saxton Community Library,

Front Street, Saxton, Pennsylvania

16678.

Attorney for the Licensee: Ernest L.

Blake, Jr., Esquire, Shaw, Pittman, Potts,

and Trowbridge, 2300 N Street, N.W.,

Washington, D.C. 20037.

NRC Project Director: Seymour H.

Weiss.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50–498 and 50–499, South Texas

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: January

28, 1997.

Description of amendment request:

The proposed amendment would

relocate the details of Technical

Specification (TS) Section 6.2.3 on the

Independent Safety Engineering Group

(ISEG) from the Administration Controls

section of the TSs and place these

details in the Updated Final Safety

Analysis Report (UFSAR) for South

Texas Project, Units 1 and 2. This

relocation is administrative only, and

would not render any changes to the

existing plant philosophy toward the

ISEG or any safety analysis. Section

6.2.3 would be deleted from the TSs and

removed from the table of contents for

Administrative Controls. Currently

UFSAR Section 13.4.2.2 describes the

ISEG, but not in the detail as the current

TSs.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes move details from

the Technical Specifications [TSs] to the

Updated Final Safety Analysis Report

(UFSAR). The changes do not result in any

hardware or operating procedure changes.

The details being removed from the

Technical Specifications [TSs] are not

assumed to be an initiator of any analyzed

event. The UFSAR, which will contain the

removed Technical Specification [TS] details,

will be maintained using the provisions of 10

CFR 50.59 and is subject to the change

control process in the Administrative

Controls Section of the Technical

Specifications [TSs]. [In addition] any

changes to the UFSAR will be evaluated per

10 CFR 50.59, no increase in the probability

or consequences of an accident previously

evaluated will be allowed without prior NRC

[Nuclear Regulatory Commission] approval.

Therefore, the changes do not involve a

significant increase in the probability or

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 move details from

the technical Specifications [TSs] to the

Updated Final Safety Analysis Report

(UFSAR). The changes will not alter the plant

configuration (no new or different type of

equipment will be installed) or make changes

in methods governing plant operation. The

changes will not impose different

requirements, and adequate control of

information will be maintained. The changes

will not alter assumptions made in the safety

analysis and licensing basis. Therefore, the

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 changes move detail from

the Technical Specifications [TSs] to the

Updated Final Safety Analysis Report

(UFSAR). The changes do not reduce the

margin of safety since the relocation of

details [is an administrative action and] has

no impact on any safety analysis

assumptions. In addition, the detail

transposed from the Technical Specifications

[TSs] to the UFSAR are the same as the

existing Technical Specification [TS] [6.2.3].

[In addition] any future changes to the FSAR

will be evaluated per the requirements of 10

CFR 50.59, no reduction in a margin of safety

will be allowed without prior NRC approval.

[Therefore, the licensee concluded that the
11496 Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices

changes will not involve a significant

reduction in a margin of safety.]

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the standards of

10 CFR 50.92(c) are satisfied. Therefore,

the NRC staff proposes to determine that

the request for amendments involves no

significant hazards consideration.

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges, Learning Center,

911 Boling Highway, Wharton, TX

77488.

Attorney for licensee: Jack R.

Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W.,

Washington, DC 20036–5869.

NRC Project Director: William D.

Beckner.

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: February

18, 1997.

Description of amendment request:

The proposed amendment would

change the reactor core fuel assembly

design features requirements contained

in Technical Specification 5.3.1, Fuel

Assemblies. The proposed change

would allow for the limited replacement

of failed or damaged fuel rods in fuel

assemblies with solid stainless steel or

zirconium alloy filler rods in

accordance with NRC-approved

applications of fuel rod configurations.

Reconstituted fuel assemblies would be

limited to those fuel designs that have

been analyzed with applicable NRCstaff-approved codes and methods and

shown by tests or analyses to comply

with all fuel safety design bases. A

limited number of lead test assemblies

that have not completed representative

testing would be allowed to be placed

in nonlimiting core regions.

The proposed change would be in

accordance with the guidance provided

in NRC Generic Letter 90–02,

Supplement 1, issued July 31, 1992.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). The

NRC staff’s review is presented below.

A. The changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated (10 CFR 50.92(c)(1)) because

the fuel assemblies would continue to

meet the same fuel assembly and fuel

rod design bases as the current fuel

assemblies, the acceptance criteria for

emergency core cooling systems would

continue to be satisfied for all fuel

assemblies, there would be no changes

to reload design and safety analysis

limits, and the radiological

consequences of accidents previously

evaluated would remain valid.

B. The changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated (10 CFR 50.92(c)(2)) because

the fuel assemblies would continue to

satisfy the same design bases previously

used. Since the original design criteria

would be met, no new accident

initiators would be introduced. All

design and performance criteria would

continue to be met for the use of

reconstituted assemblies containing the

approved filler rods. Furthermore, the

use of reconstituted fuel assemblies

does not affect the manner by which the

facility is operated.

C. The changes do not involve a

significant reduction in a margin of

safety (10 CFR 50.92(c)(3)) because the

core reload design and safety analysis

limits would be unchanged by the use

of fuel assemblies containing approved

filler rods. The use of all fuel assemblies

would continue to be limited by the

normal core operating conditions

defined in the Technical Specifications.

Reconstituted fuel assemblies would be

evaluated specifically for each cycle

reload core using approved reload

design methods and approved fuel rod

design models and methods.

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: Exeter Public Library,

Founders Park, Exeter, NH 03833.

Attorney for licensee: Lillian M.

Cuoco, Esquire, Northeast Utilities

Service Company, Post Office Box 270,

Hartford CT 06141–0270.

NRC Project Director: Patrick D.

Milano.

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: March 4,

1996.

Description of amendment request:

The proposed amendment would

modify Surveillance Requirements

4.8.1.1.2.a.6, 4.8.1.1.2.b, and

4.8.1.1.2.g.7 by specifying load bands in

loading the diesel generator (DG) in lieu

of the present requirement to load the

DG greater than or equal to a given

value. A footnote is being added to the

three surveillance requirements to

indicate that a momentary transient

outside the load range shall not

invalidate the test. The associated Bases

sections have been revised to reflect the

above changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

NNECO has reviewed the proposed

changes in accordance with 10 CFR 50.92

and has concluded that the changes do not

involve a significant hazards consideration

(SHC). The basis for this conclusion is that

the three criteria of 10 CFR 50.92(c) are not

compromised. The proposed changes do not

involve an SHC because the changes would

not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The purpose of the proposed changes to

Surveillance Requirements 4.8.1.1.2.a.6,

4.8.1.1.2.b, and 4.8.1.1.2.g.7 is to provide the

load bands for loading the DG during the

monthly, 184 days and 18-month

surveillances. Specifically, for monthly

(Surveillance 4.8.1.1.2.a.6) and once per 184

days (Surveillance 4.8.1.1.2.b) surveillances,

the load band is between 4800–5000 kW. For

the 18-month surveillance (Surveillance

4.8.1.1.2.g.7), the load band is between 5400–

5500 kW during the first 2 hours and

between 4800–5000 kW during the remaining

22 hours. The specified load bands account

for instrumentation inaccuracies using the

plant computer and for the operational

control capabilities and human factor

characteristics. The proposed changes will

keep the actual upper load limit of the DG

below the manufacturer’s recommended limit

and the actual lower limit enveloping the

accident load requirements. The proposed

changes will reduce unnecessary engine

stress and wear, while potentially improving

overall diesel generator reliability and

availability. The changes to the Bases section

reflect the changes made to the surveillance

requirements and, therefore, have no adverse

impact on plant safety. Since the proposed

changes serve to enhance overall safety, these

changes do 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.

The proposed changes regarding the load

band for the DGs do not affect the operation

or response of any plant equipment,

including the DG, or introduce any new

failure mechanism. The proposed changes

will reduce unnecessary engine stress and

wear, while potentially improving overall DG

reliability and availability. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.
Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices 11497

3. Involve a significant reduction in a

margin of safety.

The proposed changes specifying the load

bands for diesel testing will keep the actual

upper load limit of the DG below the

manufacturer’s recommended limit, and the

actual lower limit enveloping the accident

load requirements. Therefore, the proposed

changes do not affect the capability of the

diesel to perform its intended function. The

purpose of these changes is to increase the

overall DG reliability. The proposed changes

do not impact the consequences of any

design basis accidents. There is no direct

impact on any of the protective boundaries.

For these reasons, the changes do not involve

a reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

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, CT 06141–0270.

NRC Deputy Director: Phillip F.

McKee.

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: January

31, 1997.

Description of amendment request:

The amendments would revise

Technical Specification 3/4.6.1.5, and

its associated Bases section, to ensure

that a representative average

containment air temperature is

measured.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

Limitations on containment average air

temperature ensure that the overall

containment average air temperature does not

exceed the initial temperature condition

assumed in the accident analysis for a Loss

of Coolant Accident or Steamline Break

inside Containment. The resulting DBA

temperature limits are used to established the

environmental qualification envelope for

safety-related electrical equipment inside

containment.

The measurement of Containment average

air temperature is a means to ensure that the

design temperature normal operating limit is

not exceeded. The probability of an accident

is not impacted by the surveillance of normal

temperature as it is a measurement which

involves permanently installed, static

equipment. The consequences of an accident

are not impacted since the method of

measurement ensures that the design basis

temperatures are maintained and the intent

of the existing surveillance specification is

not changed. The proposed change does not

impact the actual containment temperature,

but specifies an acceptably accurate method

for its determination.

Therefore, the probability of 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

within Containment, do not introduce any

new operating configurations or new modes

of plant operation, nor change the safety

analyses. The proposed change is consistent

with NUREG–1431 and provides a

methodology to ensure that calculated

temperature is accurately determined.

The proposed changes will, therefore, 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 results in an

acceptably accurate determination of the

containment average air temperature,

therefore, compliance with the TS

surveillance and its associated basis is

assured. The present margin of safety is not

affected since operating parameters and

conditions are unchanged.

All changes are consistent with the intent

of Salem’s current TS and with the

surveillance specified in NUREG–1431,

Revision 1.

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: Mark J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW, Washington,

DC 20005–3502.

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: February

11, 1997.

Description of amendment request:

The amendments would add a new

Technical Specification 3/4.7.10,

‘‘Chilled Water System’’ to address the

support function this system provides to

other necessary safety systems.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The Chilled Water System is a support

system providing cooling to the Relay Rooms,

the Control Room, and the affected Electrical

Equipment Rooms. The Chilled Water

System is not an accident initiator of any

accident evaluated in the Safety Analysis

Report. No physical changes to the Chilled

Water System result from the proposed TS.

The specified Allowed Outage Times in the

TS are commensurate with the safety

significance of the Chilled Water System as

demonstrated by the PSA analysis.

Therefore, the proposed TS does not

significantly increase the probability or

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 involve any

modifications to the Chilled Water System or

mode of operation of the system. The

proposed TS specifies the minimum operable

number of chillers and chilled water pumps

to assure that the system performs its design

function. It does not change the basic way in

which the Chilled Water System is operated.

The loads that are isolated are non-safety

loads. By maintaining the minimum operable

number of chillers and chilled water pumps,

adequate cooling is assured to the Relay

Rooms, the Control Room, the affected

Electrical Equipment Rooms.

Therefore, the change will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The Chilled Water System is a support

system which provides cooling to the Relay

Rooms, the Control Room, and the affected

Electrical Equipment Rooms. The proposed

changes do not involve any modifications to

the Chilled Water System or changes to the

mode of operation of the system. The

proposed TS establishes controls to better

ensure that the Chilled Water System will be

able to perform its intended design function
11498 Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices

and ensures that the safety functions of

supported systems are maintained.

The proposed changes establish Allowed

Outage Times and do not affect the operation

of the Chilled Water System, and thus do 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: Salem Free Public library, 112

West Broadway, Salem, NJ 08079.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW, Washington,

DC 20005–3502.

NRC Project Director: John F. Stolz.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of amendment request: January

20, 1997.

Description of amendment request:

The proposed amendment would

change Technical Specification (TS)

Section 3/4.5.2, ‘‘Emergency Core

Cooling Systems, ECCS Subsystems—T

avg ≥ 280 °F,’’ TS Section 3/4.5.3,

‘‘Emergency Core Cooling Systems,

ECCS Subsystems—Tavg < 280°F,’’ and

TS Section 3/4.7, ‘‘Plant Systems.’’

Several surveillance intervals would be

changed from 18 months to once each

refueling interval.

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:

Toledo Edison has reviewed the proposed

changes and determined that a significant

hazards consideration does not exist because

operation of the Davis-Besse Nuclear Power

Station, Unit No. 1, in accordance with these

changes would:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because no such accidents are

affected by the proposed revisions to increase

the surveillance test intervals from 18 to 24

months for the ECCS Subsystems

(Surveillance Requirements 4.5.2.d.2.a,

4.5.2.e, 4.5.2.g.2, and 4.5.3), Auxiliary

Feedwater System (Surveillance Requirement

4.7.1.2.1.c), Motor Driven Feedwater Pump

System (Surveillance Requirement 4.7.1.7.d),

Component Cooling Water System

(Surveillance Requirement 4.7.3.1.b) and

Service Water System (Surveillance

Requirement 4.7.4.1.b). Initiating conditions

and assumptions remain as previously

analyzed for accidents in the DBNPS

Updated Safety Analysis Report.

These revisions do not involve any

physical changes to systems or components,

nor do they alter the typical manner in which

the systems or components are operated.

A review of historical 18 month

surveillance data and maintenance records

support an increase in the surveillance test

intervals from 18 to 24 months (and up to 30

months on a non-routine basis) because no

potential for a significant increase in a failure

rate of an affected system or component was

identified during these reviews.

These proposed revisions are consistent

with the NRC guidance on evaluating and

proposing such revisions as provided in

Generic Letter 91–04, ‘‘Changes in Technical

Specification Surveillance Intervals to

Accommodate a 24-Month Fuel Cycle,’’ dated

April 2, 1991.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the source term,

containment isolation or radiological releases

are not being changed by these proposed

revisions. Existing system and component

redundancy is not being changed by these

proposed changes. Existing system and

component operation is not being changed by

these proposed changes. The assumptions

used in evaluating the radiological

consequences in the DBNPS Updated Safety

Analysis Report are not invalidated.

A review of historical 18 month

surveillance data and maintenance records

support an increase in the surveillance test

intervals from 18 to 24 months (and up to 30

months on a non-routine basis) because no

potential for a significant increase in a failure

rate of an affected system or component was

identified during these reviews.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because these revisions

do not involve any physical changes to

systems or components, nor do they alter the

typical manner in which the systems or

components are operated. A review of

historical 18 month surveillance data and

maintenance records support an increase in

the surveillance test intervals from 18 to 24

months (and up to 30 months on a nonroutine basis) because no potential for a

significant increase in a failure rate of a

system or component was identified during

these reviews. No changes are being

proposed to the type of testing currently

being performed, only to the length of the

surveillance test interval.

3. Not involve a significant reduction in a

margin of safety because a review of the

historical 18 month surveillance data and

maintenance records identified no potential

for a significant increase in a failure rate of

a system or component due to increasing the

surveillance test interval to 24 months.

Existing system and component redundancy

is not being changed by these proposed

changes.

There are no new or significant changes to

the initial conditions contributing to accident

severity or consequences, therefore, there are

no significant reductions in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Gail H. Marcus.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio

Date of amendment request: January

30, 1997.

Description of amendment request:

The proposed amendment would

change Technical Specification (TS)

Section 2.2, ‘‘Limiting Safety System

Settings,’’ and applicable bases, TS

Section 3/4.3, ‘‘Instrumentation,’’ and

applicable bases, TS Section 3/4.4,

‘‘Reactor Coolant System,’’ and TS

Section 3/4.7, ‘‘Plant Systems.’’ Several

surveillance intervals would be changed

from 18 months to once each refueling

interval. In addition, several setpoints

would be revised based on an

instrument drift study, and trip

setpoints would be revised based on

new calculations. Administrative

revisions are also proposed consistent

with these changes.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

Toledo Edison has reviewed the proposed

changes and determined that a significant

hazards consideration does not exist because

operation of the Davis-Besse Nuclear Power

Station, Unit No. 1, in accordance with these

changes would:

1a. Not involve a significant increase in the

probability of an accident previously

evaluated because no such accidents are

affected by the proposed revisions to increase

the surveillance test intervals from 18 to 24

months for the subject Technical

Specifications (TS): TS 2.2 Limiting Safety

System Settings; TS 3/4.3.1.1, Reactor

Protection System Instrumentation; TS 3/

4.3.2.2, Steam and Feedwater Rupture

Control System Instrumentation; TS 3/

4.3.3.5.1, Remote Shutdown Instrumentation;
Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices 11499

TS 3/4.3.3.6, Post-Accident Monitoring

Instrumentation; TS 3/4.4.3, Safety Valves

and Pilot Operated Relief Valve—Operating;

TS 3/4.4.6.1, Reactor Coolant System Leakage

Detection Systems; TS 3/4.7.1.2 and

Auxiliary Feedwater System. Initiating

conditions and assumptions remain as

previously analyzed for accidents in the

DBNPS Updated Safety Analysis Report.

Results of the instrument drift study

analysis and review of historical 18 month

surveillance data and maintenance records

support an increase in the surveillance test

intervals from 18 to 24 months (and up to 30

months on a non-routine basis) because: the

projected instrument errors caused by drift

are bounded by the existing setpoint analysis

or either a new analysis has been performed

incorporating a more conservative setpoint or

the calculations excess margin was reduced;

projected instrument errors caused by drift

are acceptable for control of plant parameters

to effect a safe shutdown with the associated

instrumentation or an engineering evaluation

has been performed to justify continued use

of the instrument string and revisions will be

made to DBNPS calculations and controlling

procedures where appropriate, to offset any

adverse effect; and no potential for a

significant increase in a failure rate of a

system or component was identified during

surveillance data and maintenance records

reviews.

These proposed revisions are consistent

with the NRC guidance on evaluating and

proposing such revisions as provided in

Generic Letter 91–04, ‘‘Changes in Technical

Specification Surveillance Intervals to

Accommodate a 24-Month Fuel Cycle,’’ dated

April 2, 1991.

The proposed revisions to Allowable

Values for Steam and Feedwater Rupture

Control System Steam Generator Level—Low

are conservative with respect to the current

Allowable Values and therefore, do not

adversely affect previously analyzed

accidents.

The application of the Allowable Value to

the Channel Functional Test only, the

proposed deletion of the Trip Setpoint, and

revision of the Limiting Condition for

Operation and Action Statement A for TS

3.3.2.2, SFRCS Instrumentation, associated

with the proposed revision of the Allowable

Values for SFRCS Steam Generator Level—

Low are consistent with NUREG–1430,

Revision 1, ‘‘Standard Technical

Specifications, Babcock and Wilcox Plants,’’

dated April, 1995. The proposed revisions

will have no adverse effect on any previously

analyzed accident.

The proposed revision to the Reactor

Protection System High Flux Allowable

Value was determined in accordance with

the approved setpoint methodology

described in Babcock and Wilcox document

BAW–10179P, Safety Criteria for Acceptable

Cycle Reload Analyses, and is bounded by

the High Flux trip of 112% rated power

assumed in the DBNPS accident analysis.

The proposed deletion of the Trip

Setpoints, deletion of the Allowable Values

applicable to the Channel Calibration for RC

low pressure, and RC high pressure

functional units, application of Allowable

Values to the Channel Functional Test as

opposed to the Channel Calibration, and

deletion of the ‘‘**’’ and ‘‘#’’ footnotes for

Technical Specification Table 2.2–1, Reactor

Protection System Instrumentation Trip

Setpoints, and the proposed revision to TS

2.2, Limiting Safety System Settings, are

consistent with NUREG–1430, Revision 1,

‘‘Standard Technical Specifications, Babcock

and Wilcox Plants,’’ dated April, 1995. The

proposed revisions have no adverse effect on

any previously analyzed accident.

The proposed revision to Technical

Specification Table 4.3–10, Post-Accident

Monitoring Instrumentation Surveillance

Requirements, Instrument 6, Containment

Vessel Post-Accident Radiation separates the

radiation monitors to reflect the revision to

24 month surveillance intervals for the High

Range Radiation Monitors and that the

Containment Wide Range Noble Gas

monitors will remain on a 18 month

surveillance frequency is an administrative

change and does not affect previously

analyzed accidents.

The proposed revision to the Technical

Specification Bases 2.2.1, Reactor Protection

System Instrumentation Setpoints, and Bases

3/4.3.1 and 3/4.3.2, Reactor Protection

System and Safety System Instrumentation,

are administrative and do not affect

previously analyzed accidents.

Initiating conditions and assumptions

remain as previously analyzed for accidents

in the DBNPS Updated Safety Analysis

Report.

These revisions do not involve any

physical changes to systems or components,

nor do they alter the typical manner in which

the systems or components are operated.

1b. Not involve a significant increase in the

consequences of an accident previously

evaluated because the source term,

containment isolation or radiological releases

are not being changed by these proposed

revisions. Existing system and component

redundancy is not being changed by these

proposed changes. Existing system and

component operation is not being changed by

these proposed changes and the assumptions

used in evaluating the radiological

consequences in the DBNPS Updated Safety

Analysis Report are not invalidated.

2. Not create the possibility of a new or

different kind of accident from any accident

previously evaluated because these proposed

revisions do not involve any physical

changes to systems or components, nor do

they alter the typical manner in which the

systems or components are operated.

No changes are being proposed to the type

of testing currently being performed, only to

the length of the surveillance test interval.

Results of the instrument drift study

analysis and review of historical 18 month

surveillance data and maintenance records

support an increase in the surveillance test

intervals from 18 to 24 months (and up to 30

months on a non-routine basis) because: the

projected instrument errors caused by drift

are bounded by the existing setpoint analysis

or either a new analysis has been performed

incorporating a more conservative setpoint or

the calculations excess margin was reduced;

projected instrument errors caused by drift

are acceptable for control of plant parameters

to effect a safe shutdown with the associated

instrumentation or an engineering evaluation

has been performed to justify continued use

of the instrument string and revisions will be

made to DBNPS calculations and controlling

procedures where appropriate, to offset any

adverse effect; and no potential for a

significant increase in a failure rate of a

system or component was identified during

surveillance data and maintenance records

reviews.

The proposed revisions to Allowable

Values for Steam and Feedwater Rupture

Control System Steam Generator Level—Low

are conservative with respect to the current

Allowable Values and do not alter any testing

currently being performed.

The application of the Allowable Value to

the Channel Functional Test only, the

proposed deletion of the Trip Setpoint, and

revision of the Limiting Condition for

Operation and Action Statement A for TS

3.3.2.2, SFRCS Instrumentation, associated

with the proposed revision to the Allowable

Values for SFRCS Steam Generator Level—

Low are consistent with NUREG–1430,

Revision 1, ‘‘Standard Technical

Specifications, Babcock and Wilcox Plants,’’

dated April, 1995. The proposed revisions do

not alter any testing currently being

performed.

The proposed deletion of the Trip

Setpoints, deletion of the Allowable Values

applicable to the Channel Calibration for RC

lowpressure, and RC high pressure functional

units, application of Allowable Values to the

Channel Functional Test as opposed to the

Channel Calibration, and deletion of the ‘‘**’’

and ‘‘ι’’ footnotes for Technical Specification

Table 2.2–1, Reactor Protection System

Instrumentation Trip Setpoints, and the

proposed revision to TS 2.2, Limiting Safety

System Settings, are consistent with NUREG–

1430, Revision 1, ‘‘Standard Technical

Specifications, Babcock and Wilcox Plants,’’

dated April, 1995. The proposed revisions do

not alter any testing currently being

performed.

The proposed revision to the Reactor

Protection System High Flux Allowable

Value was determined in accordance with

the approved setpoint methodology

described in Babcock and Wilcox document

BAW–10179P, Safety Criteria for Acceptable

Cycle Reload Analyses, and is bounded by

the High Flux trip of 112% rated power

assumed in the DBNPS accident analysis and

does not alter any testing currently being

performed.

The proposed revision to Technical

Specification Table 4.3–10, Post-Accident

Monitoring Instrumentation Surveillance

Requirements, Instrument 6, Containment

Vessel Post-Accident Radiation separates the

radiation monitors to reflect the revision to

24 month surveillance intervals for the High

Range Radiation Monitors and that the

Containment Wide Range Noble Gas

monitors will remain on a 18 month

surveillance frequency is an administrative

change and does not alter any testing

currently being performed.

The proposed revision to the Technical

Specification Bases 2.2.1, Reactor Protection

System Instrumentation Setpoints, and Bases

3/4.3.1 and 3/4.3.2, Reactor Protection

System and Safety System Instrumentation,
11500 Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices

are administrative and do not alter any

testing currently being performed.

3. Not involve a significant reduction in a

margin of safety because The results of the

instrument drift study analysis and review of

historical 18 month surveillance data and

maintenance records support an increase in

the surveillance test intervals from 18 to 24

months (and up to 30 months on a nonroutine basis) because: the projected

instrument errors caused by drift are

bounded by the existing setpoint analysis or

either a new analysis has been performed

incorporating a more conservative setpoint or

the calculations excess margin was reduced;

projected instrument errors caused by drift

are acceptable for control of plant parameters

to effect a safe shutdown with the associated

instrumentation or an engineering evaluation

has been performed to justify continued use

of the instrument string and revisions will be

made to DBNPS calculations and controlling

procedures where appropriate, to offset any

adverse effect; and no potential for a

significant increase in a failure rate of a

system or component was identified during

surveillance data and maintenance records

reviews. Existing system and component

redundancy is not being changed by these

proposed changes.

There are no new or significant changes to

the initial conditions contributing to accident

severity or consequences, consequently there

are no significant reductions in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Gail H. Marcus.

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.

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

455, Byron Station, Unit Nos. 1 and 2,

Ogle County, Illinois

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1 and

2, Will County, Illinois

Date of amendment request:

December 21, 1995, as supplemented on

October 24, 1996.

Description of amendment request:

The proposed amendments would

relocate certain cycle-specific parameter

limits from the Technical Specifications

to the Operating Limits Report (ORL).

Date of publication of individual

notice in Federal Register: February 20,

1997 (62 FR 7804).

Expiration date of individual notice:

March 24, 1997.

Local Public Document Room

location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O.

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

455, Byron Station, Unit Nos. 1 and 2,

Ogle County, Illinois

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1 and

2, Will County, Illinois

Date of amendment request:

November 5, 1996.

Description of amendment request:

The proposed amendments would

revise the technical specifications to

allow ComEd to take credit, on a

temporary basis, for soluble boron in the

spent fuel storage water in maintaining

an acceptable margin of subcriticality.

Date of publication of individual

notice in Federal Register: February 10,

1997 (62 FR 6016).

Expiration date of individual notice:

March 12, 1997.

Local Public Document Room

location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O.

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Commonwealth Edison Company,

Docket Nos. 50–237 and 50–249,

Dresden Nuclear Power Station, Units 1

and 2, Grundy County, Illinois

Date of amendment request: February

17, 1997.

Description of amendment request:

The amendments would increase the

maximum allowable water temperature

for the Containment Cooling Service

Water inlet and the Suppression Pool.

Date of publication of individual

notice in Federal Register: February 27,

1997 (62 FR 8998).

Expiration date of individual notice:

March 31, 1997.

Local Public Document Room

location: Morris Area Public Library

District, 604 Liberty Street, Morris,

Illinois 60450.

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
Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices 11501

Room, the Gelman Building, 2120 L

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

local public document rooms for the

particular facilities involved.

Baltimore Gas and Electric Company,

Docket No. 50–317, Calvert Cliffs

Nuclear Power Plant, Unit No. 1, Calvert

County, Maryland

Date of application for amendment:

October 3, 1996.

Brief description of amendment: The

amendment concerns the provisions at

Calvert Cliffs Unit 1 for receiving,

possessing, and using byproduct,

source, and special nuclear material.

The amendment changed the Unit 1

license, which previously contained

restrictions on the possession and use of

byproduct, source, or special nuclear

material, to be consistent with the Unit

2 license, which has no such

restrictions. The staff found this license

amendment to be acceptable since both

units share the same radiation

protection staff, and the training and

procedures used to control the

acceptance and use of radioactive

material at Unit 2 are sufficient to

control the radioactive material at Unit

1, as well.

Date of issuance: February 19, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 220.

Facility Operating License No. DPR–

53: Amendment revised the Operating

License.

Date of initial notice in Federal

Register: November 6, 1996 (61 FR

57482). The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

February 19, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.

Duke Power Company, Docket Nos. 50–

269, 50–270, and 50–287, Oconee

Nuclear Station, Units 1, 2, and 3,

Oconee County, South Carolina

Date of application for amendments:

February 20, 1996 as supplemented

October 16, 1996.

Brief description of amendments: The

amendments revise Technical

Specification (TS) 3.1.5, TS 3.1.10 and

TS 4.1 to: (1) reduce the surveillance

frequency for the boron concentration in

the concentrated boric acid storage tank;

(2) delete the surveillance requirements

for Sr89 and Sr90, gross beta activity,

gross alpha activity and dissolved gas

concentration in the reactor coolant, and

gross beta activity in the steam generator

feedwater; (3) relocate the surveillance

requirements for tritium, chloride,

fluoride, and oxygen in the reactor

coolant to the Selected Licensee

Commitment (SLC) manual; and (4)

delete TS 3.1.10 related to temperature

and pressure requirements to avoid gas

bubble formation on depressurization.

Date of issuance: February 19, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days. Implementation shall include

concurrent revision of the Selected

Licensee Commitment Manual in

accordance with the application of this

amendment.

Amendment Nos.: 221, 221, 218.

Facility Operating License Nos. DPR–

38, DPR–47 and DPR–55: Amendments

revise the Technical Specifications.

Date of initial notice in Federal

Register: March 27, 1996 (61 FR 13523).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated February 19, 1997.

No significant hazards consideration

comments received: No

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina.

Indiana Michigan Power Company,

Docket Nos. 50–315 and 50–316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of application for amendments:

February 26, 1996.

Brief description of amendments: The

amendments revise the TS to allow an

increased limit for the nominal

enrichment of new (unirradiated)

Westinghouse-fabricated fuel stored in

the new fuel storage racks.

Date of issuance: February 27, 1997.

Effective date: February 27, 1997,

with full implementation within 45

days.

Amendment Nos.: 213 and 198.

Facility Operating License Nos. DPR–

58 and DPR–74. Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: April 24, 1996 (61 FR 18172)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated February 27, 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.

North Atlantic Energy Service

Corporation, Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: June 4,

1996 as supplemented by letter dated

January 8, 1997.

Description of amendment request:

The amendment revises Seabrook

Appendix A Technical Specifications

(TS) 1.7, ‘‘Containment Integrity’’, 3/

4.6.1, ‘‘Primary Containment’’, and 3/

4.6.5, ‘‘Containment Enclosure

Building’’, to incorporate the provisions

of Option B to 10 CFR Part 50,

Appendix J. TS Section 6.15,

‘‘Containment Leakage Rate Testing

Program’’, has been added to establish a

Containment Leakage Rate Testing

Program, as specified in Regulatory

Guide 1.163, dated September 1995, to

support these changes. In addition to

the changes to incorporate the

provisions of Option B, TS 3.6.1.7 and

4.6.1.7.1 have been revised to

incorporate an increased leak testing

interval and to include reference to the

Containment Leakage Rate Testing

Program.

Date of issuance: February 24, 1997.

Effective date: February 24, 1997.

Amendment No.: 49.

Facility Operating License No. NPF–

86. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 28, 1996 (61 FR

44359). The licensee’s letter dated

January 8, 1997, which provided

additional information relating to

containment purge supply and exhaust

valve testing and maintenance, does 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 February 24,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

Northeast Nuclear Energy Company, et

al., Docket No. 50–423, Millstone

Nuclear Power Station, Unit No. 3, New

London County, Connecticut

Date of application for amendment:

July 18, 1995.

Brief description of amendment: The

amendment revises the Technical

Specifications (TS) to extend the

surveillance schedule from 18 months

to each refueling interval (nominally 24

months) for TS 3/4.4.4, ‘‘Relief Valves;’’

TS 3/4.4.6.1, ‘‘Reactor Coolant System
11502 Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices

Leakage;’’ TS 3/4.4.6.2, ‘‘Operational

Leakage;’’ TS 3/4.4.9.3, ‘‘Overpressure

Protection Systems;’’ and TS 3/4.4.11,

‘‘Reactor Coolant System Vents.’’

Date of issuance: February 19, 1997.

Effective date: As of the date of

issuance, to be implemented within 90

days.

Amendment No.: 133.

Facility Operating License No. NPF–

49: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 27, 1995 (60 FR

58402).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 19,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut 06360, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut 06385

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:

October 25, 1996.

Brief description of amendments: The

amendments revise the Technical

Specifications (TSs) to incorporate the

requirements of 10 CFR Part 50,

Appendix J, Option B, for containment

leakage tests. In addition, the

amendments add a new section to the

TSs, which establishes the requirements

of the containment leakage rate testing

program, consistent with the Improved

Standard Technical Specifications.

Date of issuance: February 19, 1997.

Effective date: February 19, 1997,

with full implementation within 30

days.

Amendment Nos.: 126 and 118.

Facility Operating License Nos. DPR–

42 and DPR–60. Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: January 15, 1997 (62 FR 2191)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated February 19, 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.

Omaha Public Power District, Docket

No. 50–285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request:

November 16, 1995, as supplemented by

letter dated August 8, 1996.

Brief description of amendment: The

amendment revises the technical

specifications to add a limiting

condition for operation and surveillance

test for safety related inverters and

deletes the nonsafety related instrument

buses.

Date of issuance: February 13, 1997.

Effective date: February 13, 1997, to

be implemented within 60 days from

the date of issuance.

Amendment No.: 180.

Facility Operating License No. DPR–

40. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 13, 1996 (61 FR 10395)

The August 8, 1996, supplemental

letter provided additional clarifying

information and did not change the

initial no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 13,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102.

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:

August 27, 1996.

Brief description of amendments: The

proposed amendments change the

minimum allowable charging water

header pressure from a value of 955 psig

to a value of 940 psig in Technical

Specification 3.10.8, ‘‘Shutdown Margin

(SDM) Test-Refueling.’’

Date of issuance: February 19, 1997.

Effective date: Both units, as of date

of issuance, to be implemented within

30 days.

Amendments Nos.: 218 and 221.

Facility Operating License Nos. DPR–

44 and DPR–56: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: October 23, 1996 (61 FR

55036)

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated February 19,

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.

Pennsylvania Power and Light

Company, Docket Nos. 50–387 and 50–

388 Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments:

February 2, 1996, as supplemented

September 23, 1996.

Brief description of amendments:

These amendments change Technical

Specification 3.6.1.2 for each unit to

permit primary containment leakage

testing of the main steamline isolation

valves at either 22.5 psig or 45 psig

according to the type of test to be

conducted.

Date of issuance: February 25, 1997.

Effective date: Both units, as of date

of issuance, to be implemented within

30 days.

Amendment Nos.: 163 and 134.

Facility Operating License Nos. NPF–

14 and NPF–22. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: August 14, 1996 (61 FR

42282). The September 23, 1996, letter

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated February 25,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701.

Southern California Edison Company, et

al., Docket No. 50–362, San Onofre

Nuclear Generating Station, Unit No. 3,

San Diego County, California

Date of application for amendment:

January 14, 1997.

Brief description of amendment: The

amendment revises Surveillance

Requirements (SRs) 3.8.1.14 and

3.8.1.15 to temporarily restore

provisions of the emergency diesel

generator surveillance requirements as

they were prior to their revision as part

of NRC Amendment No. 116

(conversion to the Improved Technical

Specifications).
Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices 11503

Date of issuance: February 10, 1997.

Effective date: February 10, 1997.

Amendment Nos.: 125.

Facility Operating License Nos. NPF–

15: The amendments revised the

Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: Yes (62 FR 3536 dated

January 23, 1997). The notice provided

an opportunity to submit comments on

the Commission’s proposed no

significant hazards consideration

determination. No comments have been

received. The notice also provided for

an opportunity to request a hearing by

February 24, 1997, but indicated that if

the Commission makes a final no

significant hazards consideration

determination any such hearing would

take place after issuance of the

amendment.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 10,

1997.

Attorney for licensee: T. E. Oubre,

Esquire, Southern California Edison

Company, P. O. Box 800, Rosemead,

California 91770.

Local Public Document Room

location: Main Library, University of

California, P. O. Box 19557, Irvine,

California 92713.

Wisconsin Electric Power Company,

Docket Nos. 50–266 and 50–301, Point

Beach Nuclear Plant, Unit Nos. 1 and 2,

Town of Two Creeks, Manitowoc

County, Wisconsin

Date of application for amendments:

September 19, 1996, supplemented on

November 18, 1996, revised on January

13, 1997, and supplemented on January

27, 1997.

Brief description of amendments:

These amendments revise the reactor

coolant system temperature below

which the low temperature overpressure

protection (LTOP) system and

pressurizer power-operated relief valves

(PORVs) shall be operable, modify the

requirement to limit operation of the

high pressure safety injection pump

from reactor coolant system cold leg

temperature of less than or equal to 275

°F to whenever the LTOP is required to

be operable, change the name of the

system from the overpressure mitigation

system to the LTOP system, and revise

the PORV setpoint from 425 psig to 440

psig.

Date of issuance: February 20, 1997,

with full implementation within 45

days.

Effective date: February 20, 1997.

Amendment Nos.: 172 and 176.

Facility Operating License Nos. DPR–

24 and DPR–27: Amendments revised

the Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration (NSHC): Yes (62 FR 5256,

dated February 4, 1997) The notice

provided an opportunity to submit

comments on the Commission’s

proposed NSHC determination. No

comments have been received. The

notice also provided for an opportunity

to request a hearing by March 6, 1997,

but indicated that if the Commission

makes a final NSHC determination, any

such hearing would take place after

issuance of the amendments. The

Commission’s related evaluation of the

amendments, finding of exigent

circumstances, and final determination

of no significant hazards considerations

are contained in a Safety Evaluation

dated February 20, 1997.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

Local Public Document Room

location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241.

Wolf Creek Nuclear Operating

Corporation, Docket No. 50–482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request:

December 13, 1995, as supplemented by

letter dated October 10, 1996.

Brief description of amendment: The

amendment revises the 125-volt D.C.

Sources (3.8.2.1 and 3.8.2.2) and Onsite

Power Distribution (3.8.3.1 and 3.8.3.2)

Technical Specifications to include

provisions for installed spare battery

chargers, which will be added to the

plant design before startup from the

ninth refueling outage.

Date of issuance: February 10, 1997.

Effective date: February 10, 1997, to

be implemented before startup from the

ninth refueling outage, currently

scheduled to begin in September 1997.

Amendment No.: 104.

Facility Operating License No. NPF–

42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 22, 1996 (61 FR 1639)

The October 10, 1996, supplemental

letter provided additional clarifying

information and did not change the

initial no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 10,

1997.

No significant hazards consideration

comments received: No.

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

Notice of Issuance of Amendment to

Facility Operating License and Final No

Significant Hazards Consideration

Determination

During the period since publication of

the last biweekly notice, individual

notices of issuance of amendments have

been issued for the facilities as listed

below. These notices were previously

published as separate individual

notices. They are repeated here because

this biweekly notice lists all

amendments that have been issued for

which the Commission has made a final

determination that an amendment

involves no significant hazards

consideration.

In this case, a prior Notice of

Consideration of Issuance of

Amendment, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing was

issued, a hearing was requested, and the

amendment was issued before any

hearing because the Commission made

a final determination that the

amendment involves no significant

hazards consideration.

Details are contained in the

individual notice as cited.

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
11504 Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices

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

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

April 11, 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.
Federal Register / Vol. 62, No. 48 / Wednesday, March 12, 1997 / Notices 11505

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 20555–001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

Union at 1–(800) 248–5100 (in Missouri

1–(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–001, 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).

Northern States Power Company,

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

Island Nuclear Generating Plant,

Goodhue County, Minnesota

Date of application for amendments:

February 6, 1997, as supplemented

February 12, 1997.

Brief description of amendments: The

amendments revise Technical

Specification 3.3.A to allow safety

injection pump testing and evolutions

during low-temperature shutdown

conditions provided controls for reactor

coolant system conditions are in place

to provide low temperature

overpressurization protection.

Date of issuance: February 20, 1997.

Effective date: February 20, 1997,

with full implementation within 30

days.

Amendment Nos.: 127 and 119.

Facility Operating License Nos. DPR–

42 and DPR–60. Amendments revised

the Technical Specifications and Bases.

Public comments requested as to

proposed no significant hazards

consideration (NSHC): Yes. NRC

published a public notice of the

proposed amendments, issued a

proposed finding of no significant

hazards consideration, and requested

that any comments on the proposed

finding be provided to the staff by close

of business on February 14, 1997. The

notice was published in the Red Wing

Republican Eagle on February 12, 1997,

the Minneapolis Star Tribune on

February 9, 1997, and the St. Paul

Pioneer Press on February 10, 1997. No

comments have been received.

The Commission’s related evaluation

of the amendments, finding of exigent

circumstances, consultation with the

State of Minnesota, and final

determination of NSHC are contained in

a Safety Evaluation dated February 20,

1997.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW, Washington, DC

20037.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Dated at Rockville, Maryland, this 5th day

of March 1997.

For The Nuclear Regulatory Commission.

Jack W. Roe,

Director, Division of Reactor Projects—III/IV

Office of Nuclear Reactor Regulation.

[FR Doc. 97–5999 Filed 3–11–97; 8:45 am]

BILLING CODE 7500–01–P

OFFICE OF MANAGEMENT AND

BUDGET

Interpretation Numbers 1 and 2 Related

to Statement of Federal Financial

Accounting Standards Numbers 4, 5,

and 7

AGENCY: Office of Management and

Budget.

ACTION: Notice of interpretations.

SUMMARY: This notice includes two

interpretations of Statements of Federal

Financial Accounting Standards

(SFFAS), adopted by the Office of

Management and Budget (OMB). These

interpretations were recommended by

the Federal Accounting Standards

Advisory Board (FASAB) and adopted

in their entirety by OMB.

FOR FURTHER INFORMATION CONTACT:

Norwood J. Jackson, Jr. (telephone: 202–

395–3993), Office of Federal Financial

Management, Office of Management and

Budget.

SUPPLEMENTARY INFORMATION: This

Notice includes two interpretations of

Statements of Federal Financial

Accounting Standards (SFFAS), adopted

by the Office of Management and

Budget (OMB). These interpretations

were recommended by the Federal

Accounting Standards Advisory Board

(FASAB) and adopted in their entirety

by OMB.

Under a Memorandum of

Understanding among the General

Accounting Office, the Department of

the Treasury, and OMB on Federal

Government Accounting Standards, the

Comptroller General, the Secretary of

the Treasury, and the Director of OMB

(the Principals) decide upon standards

and concepts after considering the

recommendations of FASAB. After

agreement to specific standards and

concepts, they are published in the

Federal Register and distributed

throughout the Federal Government.

An Interpretation is a document,

originally developed by FASAB, of

narrow scope which provides

clarification of the meaning of a

standard, concept or other related

guidance. Once approved by the

designated representatives of the

Principals, they are published in the

Federal Register.

This Notice, including the first two

interpretations of SFFAS, is available on

the OMB home page on the internet

which is currently located at http://

www.whitehouse.gov/WH/EOP/OMB/

html/ombhome.html, under the caption

‘‘Federal Register Submissions.’’

G. Edward DeSeve,

Controller, Office of Federal Financial

Management, Office of Management and

Budget.

Interpretation Number 1 of Statement

of Federal Financial Accounting

Standards Number 7

Reporting on Indian Trust Funds in

General Purpose Financial Reports of

the Department of the Interior (DOI) and

in the Consolidated Financial

Statements of the United States

Government: An Interpretation of

SFFAS No. 7

Introduction

1. The DOI requested guidance about

how to report information on Indian

trust funds in the general purpose

financial report of the Department. The

Indian trust funds are managed by DOI’s

Office of Special Trustee, Office of the

Secretary. (Prior to FY 1996, the trust

funds were managed by the Bureau of

Indian Affairs.) Some of the funds

belong to individual Indians, others

belong to tribes. The funds are managed

by the Federal Government in a trust

arrangement. While the government’s

responsibility for all of these funds is of

a fiduciary nature, some portion of the

annual flows for some of the funds have