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

NUCLEAR REGULATORY COMMISSION

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

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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

[[Page 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 pressure-temperature 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 pressure-temperature 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 pressure-

temperature 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 pressure-temperature 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 pressure-temperature 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 pressure-temperature 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.

[[Page 11486]]

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 one-time 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

[[Page 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-of-coolant

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 as-found

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 non-linear 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.

[[Page 11488]]

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 re-energization 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.

[[Page 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

[[Page 11490]]

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-to-secondary 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 non-conservative 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 steady-state 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.

[[Page 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 pre-and 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-High-High';

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 Diesel-Driven 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 long-term 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

[[Page 11492]]

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 high-high 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, ``Performance-Based Requirements,'' to

10 CFR 50, Appendix J, ``Primary Reactor Containment Leakage Testing

for Water-Cooled 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.

[[Page 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 performance-based 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

[[Page 11494]]

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

[[Page 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

[[Page 11496]]

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 NRC-staff-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.

[[Page 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

[[Page 11498]]

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 deg.F,''

TS Section 3/4.5.3, ``Emergency Core Cooling Systems, ECCS Subsystems--

Tavg ''

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,

[[Page 11500]]

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 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. 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

[[Page 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

[[Page 11502]]

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).

[[Page 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 deg.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

[[Page 11504]]

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.

[[Page 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 propo

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