Biweekly Notice

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory

Commission (the Commission or NRC staff) is publishing this regular

biweekly notice. Public Law 97-415 revised section 189 of the Atomic

Energy Act of 1954, as amended (the Act), to require the Commission to

publish notice of any amendments issued, or proposed to be issued,

under a new provision of section 189 of the Act. This provision grants

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from October 27, 1997, through November 6, 1997.

The last biweekly notice was published on November 5, 1997 (62 FR

59912).

Notice Of Consideration Of Issuance Of Amendments ToFacility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

the Commission's regulations in 10 CFR 50.92, this means that operation

of the facility in accordance with the proposed amendment would not (1)

involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

or (3) involve a significant reduction in a margin of safety. The basis

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

determination. Any comments received within 30 days after the date of

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

[[Page 61837]]

expiration of the 30-day notice period. However, should circumstances

change during the notice period such that failure to act in a timely

way would result, for example, in derating or shutdown of the facility,

the Commission may issue the license amendment before the expiration of

the 30-day notice period, provided that its final determination is that

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

before action is taken. Should the Commission take this action, it will

publish in the Federal Register a notice of issuance and provide for

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules and

Directives Branch, Division of Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and should cite the publication date and

page number of this Federal Register notice. Written comments may also

be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,

Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.

Copies of written comments received may be examined at the NRC Public

Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.

The filing of requests for a hearing and petitions for leave to

intervene is discussed below.

By December 19, 1997, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and any person whose interest may be affected by this

proceeding and who wishes to participate as a party in the proceeding

must file a written request for a hearing and a petition for leave to

intervene. Requests for a hearing and a petition for leave to intervene

shall be filed in accordance with the Commission's ``Rules of Practice

for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested

persons should consult a current copy of 10 CFR 2.714 which is

available at the Commission's Public Document Room, the Gelman

Building, 2120 L Street, NW., Washington, DC and at the local public

document room for the particular facility involved. If a request for a

hearing or petition for leave to intervene is filed by the above date,

the Commission or an Atomic Safety and Licensing Board, designated by

the Commission or by the Chairman of the Atomic Safety and Licensing

Board Panel, will rule on the request and/or petition; and the

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene

shall set forth with particularity the interest of the petitioner in

the proceeding, and how that interest may be affected by the results of

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

following factors: (1) the nature of the petitioner's right under the

Act to be made a party to the proceeding; (2) the nature and extent of

the petitioner's property, financial, or other interest in the

proceeding; and (3) the possible effect of any order which may be

entered in the proceeding on the petitioner's interest. The petition

should also identify the specific aspect(s) of the subject matter of

the proceeding as to which petitioner wishes to intervene. Any person

who has filed a petition for leave to intervene or who has been

admitted as a party may amend the petition without requesting leave of

the Board up to 15 days prior to the first prehearing conference

scheduled in the proceeding, but such an amended petition must satisfy

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

scheduled in the proceeding, a petitioner shall file a supplement to

the petition to intervene which must include a list of the contentions

which are sought to be litigated in the matter. Each contention must

consist of a specific statement of the issue of law or fact to be

raised or controverted. In addition, the petitioner shall provide a

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

contention and on which the petitioner intends to rely in proving the

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

aware and on which the petitioner intends to rely to establish those

facts or expert opinion. Petitioner must provide sufficient information

to show that a genuine dispute exists with the applicant on a material

issue of law or fact. Contentions shall be limited to matters within

the scope of the amendment under consideration. The contention must be

one which, if proven, would entitle the petitioner to relief. A

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses.

If a hearing is requested, the Commission will make a final

determination on the issue of no significant hazards consideration. The

final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

request for a hearing. Any hearing held would take place after issuance

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555-0001, Attention: Rulemakings and

Adjudications Staff, or may be delivered to the Commission's Public

Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,

by the above date. A copy of the petition should also be sent to the

Office of the General Counsel, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for a hearing will

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

petition and/or request should be granted based upon a balancing of

factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this action, see the

application for amendment which is available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the local public document room for

the particular facility involved.

[[Page 61838]]

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert

County, Maryland

Date of amendments request: October 22, 1997

Description of amendments request: The proposed amendment

incorporates both steady state and transient degraded voltage setpoints

into Technical Specifications, as opposed to the current single

degraded voltage setpoint. The proposed changes ensure adequate

terminal voltage to all safety-related equipment during steady state

and transient voltage conditions. Additionally, the 4 kV voltage range

required during testing of the emergency diesel generators (EDGs) will

be decreased to ensure the new steady state degraded voltage relays are

not actuated during testing and to ensure the 4 kV motors are operated

within their voltage rating.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The proposed changes revise the current degraded voltage

setpoint and adds an additional steady state undervoltage

requirement to Unit 1 and 2 Technical Specifications. The current

degraded voltage relays will be referred to as

transient degraded voltage relays.'' The new settings

allow for calibration tolerances, potential transformer correction

factors, test equipment uncertainties, and relay drift. The nominal

settings account for the above factors, plus additional margin to

the analytical limit. The acceptable voltage range during EDG

surveillance testing is also being decreased. The setpoint and time

delay associated with the 4 kV bus loss of voltage relays is

unaffected by this amendment request.

The accident analyses credit the loading of the EDGs based on

loss of offsite power. The 4 kV emergency bus loss of voltage and

degraded voltage relays initiate starting and loading of the

emergency diesel generators (EDGs) when the preferred power source

voltage is lost or drops below a predetermined value. The relays

also initiate disconnection of the preferred power source from the 4

kV emergency busses. These actions ensure adequate terminal voltage

to all safety-related electrical equipment required to support

accident mitigation. The required voltage necessary to ensure

safety-related motors are capable of starting is 75 percent of

nominal rated equipment voltage. The required voltage necessary to

ensure these motors continue running for extended periods is 90

percent of nominal rated equipment voltage.

The degraded (transient) voltage setpoint is being changed from

3628 [plus or minus] 25 Volts to 3710 [plus or minus] 80 Volts.

Based on the most recent calculations, a minimum voltage of 3630

Volts is required to ensure at least 75 percent of the nominal

voltage is available to No. 13 Charging Pump, which is the most

limiting electrical load.

The new steady state degraded voltage relay setpoint will be

established at 3900 [plus or minus] 80 Volts. The setpoint ensures

that there is at least 90 percent of nominal voltage available to

No. 13 Charging Pump. The time delay associated with this actuation

is 101 [plus or minus] 3.5 seconds. The time delay provides adequate

time for the voltage regulator to recover bus voltage following a

voltage swing on the 500 kV system and time for the EDG voltage

regulator to stabilize. The steady state degraded voltage relays

will be tested in the same manner, and at the same frequency, as the

loss of voltage and transient degraded voltage relays.

The required voltage range during EDG surveillance testing is

being revised from 4160 [plus or minus] 420 Volts to 4160 +240, -100

Volts. The surveillance requirement verifies that the EDG voltage

regulator is maintaining an acceptable voltage. The

new value ensures the 4 kV motors are operated within their

rated voltage and prevents actuation of the steady state degraded

voltage relay during surveillance testing.

The degraded voltage relays are not initiators in any previously

evaluated accidents. Additionally, decreasing the acceptable voltage

range during EDG testing does not affect the initiation of any

previously analyzed accidents. Therefore, the proposed changes do

not involve an increase in the probability or consequences of an

accident previously analyzed.

2. Would not create the possibility of a new or different type

of accident from any accident previously evaluated.

The license amendment request revises the current degraded

voltage setpoint and adds an additional steady state degraded

voltage requirement. Additionally, the acceptable voltage range

during EDG surveillance testing is being decreased. The proposed

changes ensure adequate starting and running terminal voltage to all

safety-related electrical equipment during steady state and

transient degraded voltage conditions. The addition of the steady

state degraded voltage relays provide an extra scheme of protection

against sustained degraded voltage conditions. The facility

currently relies upon degraded voltage relays to start and load the

EDGs and to disconnect the preferred power source from the 4 kV

emergency busses. Therefore, revising the relay setpoint, adding

additional steady state degraded voltage protection, and decreasing

the acceptable voltage range during EDG testing does not create the

possibility of a new or different type of accident from any accident

previously evaluated.

3. Would not involve a significant reduction in a margin of

safety.

The safety function of the degraded voltage relays is to ensure

that the preferred power source is disconnected from the 4 kV

emergency busses during loss of voltage or degraded voltage

conditions. The relays also ensure the EDGs are started and loaded.

Ultimately, these actions ensure the minimal terminal voltage

necessary to start and run all safety-related electrical equipment

is maintained. The proposed changes revise the current degraded

voltage setpoint and adds an additional steady state undervoltage

requirement. Additionally, the acceptable voltage range during EDG

surveillance testing is being decreased to ensure actuation of the

steady state degraded voltage relays does not occur during

EDG testing, and to ensure the 4 kV motors are operated within

their rated voltage range.

Because the proposed changes ultimately ensure adequate terminal

voltage to all safety-related electrical equipment during transient

and steady state undervoltage conditions, the safety function of the

degraded voltage relays, as well as the margin of safety afforded by

these relays is unchanged.

Therefore, the changes do not involve a significant reduction in

the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendments request involves no significant hazards consideration.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678

Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,

Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037

NRC Project Director: S. Singh Bajwa, Director

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson

Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: October 2, 1997

Description of amendment request: The proposed amendment would

address an unreviewed safety question associated with the analysis of a

fuel handling accident in the Fuel Storage Building as described in

Section 15.7.4, ``Design Basis Fuel Handling Accidents,'' of the H.B.

Robinson Steam Electric Plant (HBR) Updated Final Safety Analysis

Report (UFSAR). Carolina Power & Light Company (the licensee)

determined that an assumption used in the accident analysis for depth

of water above the top of irradiated fuel in the spent fuel pit was

non-conservative. The accident analysis assumed a depth of 23 feet

instead of the correct value of 21 feet. The licensee has submitted a

revised accident analysis using the correct assumption and has proposed

that the UFSAR be

[[Page 61839]]

changed to incorporate the results of the revised analysis.

Basis for proposed no significant hazards determination: As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration, which is presented

below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change to the UFSAR is to change assumptions

associated with the evaluation of a fuel handling accident in the

Fuel Storage Building. The change in assumptions is to reduce the

decontamination factor associated with the removal of elemental

iodine from the spent fuel pool water. Because the decontamination

factor for elemental iodine is reduced, the consequences of a fuel

handling accident in the Fuel Storage Building is [sic] increased.

However, because the radiological consequences remain well within

the exposure guideline values of 10 CFR 100, paragraph 11 (i.e., 25%

or less of the values), the increase in consequences is not

significant. The change in assumptions for the fuel handling

accident in the Fuel Storage Building do [sic] not affect operation,

maintenance, or design of equipment associated with the handling of

fuel in the Fuel Storage Building, therefore, the probability of a

fuel handling accident as previously evaluated is not changed.

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed change does not involve any physical alteration of

plant systems, structures or components, changes in parameters

governing normal plant operation, or methods of operation. The

proposed change does not introduce a new mode of operation or

changes in the method of normal plant operation. Therefore, the

possibility of a new or different kind of accident from any accident

previously evaluated is not created.

3. Does this change involve a significant reduction in a margin

of safety?

The proposed change to the UFSAR to change the assumptions

associated with a fuel handling accident in the Fuel Storage

Building is to change the assumption for the decontamination factor

for elemental iodine to a smaller value. The new assumption for

elemental iodine decontamination factor preserves the approximate

factor of 24 margin between experimental data for elemental iodine

decontamination factor and the assumed value provided in NRC Safety

Guide 25. Therefore, the change does not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550

Attorney for licensee: R. E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602

NRC Project Director: James E. LyonsCommonwealth Edison Company,

Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois Docket Nos. 50-254 and 50-265, Quad

Cities Nuclear Power Station, Units 1 and 2, Rock Island County,

Illinois Date of application for amendment request: September 30, 1997

Description of amendment request: This request changes the

Technical Specifications (TS) by adding a new Section 3/4.12.C,

``Inservice Leak and Hydrostatic Testing Operation,'' to allow certain

reactor coolant pressure tests to be performed in MODE 4 when the

metallurgical characteristics of the reactor pressure vessel require

the pressure testing at or approaching temperatures

212 deg.F, which normally correspond with MODE 3.

Basis for proposed no significant hazards determination: As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration, which is presented

below: 1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because of the

following:

The proposed amendment represents the addition of a Special Test

Exception to perform Pressure Testing Operations consistent with the

requirements of Section 3.10.1 of the Improved Standard Technical

Specifications (NUREG-1433). The proposed changes are consistent

with the current plant safety analyses. Implementation of these

changes will provide continued assurance that specified parameters

associated with Pressure Testing Operations will remain within their

acceptance limits, and as such, will not significantly increase the

probability or consequences of a previously evaluated accident.

The proposed changes are based on requirements specified by

Section 3.10.1 of NUREG-1433. Any such changes are consistent with

the current plant safety analyses and have been determined to

represent sufficient requirements for the assurance and reliability

of equipment assumed to operate in the safety analyses, or provide

continued assurance that specified parameters associated with

Pressure Testing Operations remain within their acceptance limits.

As such, these changes will not significantly increase the

probability or consequences of a previously evaluated accident.

The associated systems affecting Pressure Testing Operations

related to this proposed amendment are not assumed in any analyses

to initiate any accident sequence; therefore, the probability of any

accident previously evaluated is not increased by this proposed

amendment which incorporates the requirements of Section 3.10.1 of

NUREG-1433. In addition, the proposed limiting conditions for

operation and surveillance requirements for the proposed amendment

ensure a level of equipment operability sufficient to mitigate any

operational occurrences which could occur while operating under this

Special Test Exception. Furthermore, any operational occurrence

postulated during operation under this Special Test Exception is

bounded by the Design Basis Accidents. Therefore, the proposed

amendment does not increase the consequences of nay accident

previously evaluated.

There is no change to the consequences of an accident previously

evaluated because Pressure Testing Operations does not adversely

affect either the on-site or off-site does consequences resulting

from an accident. In addition, Pressure Testing Operations is not an

accident initiator. As such, there is no adverse impact on the

probability of accident initiators. Thus, there is no significant

increase in the probability of any previously analyzed accident.

2) Create the possibility of a new or different kind of accident

from any accident previously evaluated because:

The proposed amendment represents the conversion of current

Technical Specification requirements to maintain consistency with

those requirements specified in Section 3.10.1 of NUREG-1433. The

proposed changes are consistent with the current plant safety

analyses. These proposed changes do not involve revisions to the

design of the station. In addition, the proposed limiting conditions

for operation and surveillance requirements for the proposed

amendment ensure a level of equipment operability sufficient to

mitigate any operational occurrences which could occur while

operating under the Special Test Exception. Some of the changes may

involve revision in the testing of components at the station;

however, these are in accordance with the current plant safety

analyses. The proposed changes will not introduce new failure

mechanisms beyond those already considered In the current plant

safety analyses.

The associated systems that affect Pressure Testing Operations

related to the proposed amendment, are not assumed in any plant

safety analysis to initiate any accident sequence. In addition, the

proposed surveillance requirements for any such affected systems are

consistent with the requirements of Section 3.10.1 of NUREG-1433.

Therefore, the possibility of a new or different kind of accident

from any accident previously evaluated is not created.

3) Involve a significant reduction in the margin of safety

because:

ComEd proposes to revise the Technical Specifications to be

consistent with those provisions specified in Section 3.10.1 of

NUREG-1433. The proposed changes are consistent with the current

plant safety analyses. In addition, these proposed changes do not

involve revisions to the design of the station. As such, the

proposed individual changes will maintain the same level of

[[Page 61840]]

reliability of the equipment associated with Pressure Testing

Operations, assumed to operate in the plant safety analysis, or

provide continued assurance that specified parameters affecting,

will remain within their acceptance limits. Therefore, the proposed

changes provide continued assurance of Pressure Testing Operations

without adversely affecting the public health and safety and as

such, will not significantly reduce existing plant safety margins.

The proposed amendment to the Technical Specifications

implements present requirements, or the requirements in accordance

with the guidelines set forth in Section 3.10.1 of NUREG-1433. The

proposed changes have been evaluated and found to be acceptable for

use at the stations based on system design, safety analysis

requirements, and operational performance. Since the proposed

changes are based on NRC accepted provisions that are applicable at

the stations and maintain necessary levels of system or component

reliability affecting Pressure Testing Operations, the proposed

changes do not involve a significant reduction in the margin of

safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92 are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: for Dresden, Morris Area

Public Library District, 604 Liberty Street, Morris, Illinois 60450;

for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,

Illinois 61021

Attorney for licensee: Michael I. Miller, Esquire; Sidle and

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: September 26, 1997

Description of amendment request: The proposed amendments would

revise the Technical Specifications to (1) prohibit the simultaneous

opening of the drywell and suppression chamber purge system isolation

valves, (2) upgrade the ventilation filter testing program to the

latest industry standards, and (3) specify that the auxiliary electric

equipment room is required to be habitable during design bases

accidents.

Basis for proposed no significant hazards determination: As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration, which is presented

below:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

a. Drywell and Suppression Chamber Purge System

The purpose of the drywell and suppression chamber purge system

isolation valves is to mitigate the consequences of a design bases

accident. Operation of these valves will have no effect on the

probability of a design bases accident occurring.

The current TS 3.6.1.8 allows for the drywell and suppression

chamber purge system isolation valves to be open simultaneously. In

this condition, containment pressure and offsite dose during design

bases accidents would be greater than previously evaluated. The

proposed revision to TS 3.6.1.8 would prevent the simultaneous

opening of the drywell and suppression chamber purge system

isolation valves thus assuring that the consequences of design bases

accidents previously evaluated are still bounding.

b. Ventilation Filter Testing Program

The SBGTS [Standby Gas Treatment System] and Control Room and

AEER [Auxiliary Electric Equipment Room] Emergency Filtration

Systems are designed to mitigate the radiological consequences of

previously evaluated design bases accidents. Operation and testing

of these systems will have no effect on the probability of a design

bases accident occurring.

The proposed revisions associated with this change relocate the

requirements for SBGTS and Control Room and AEER Emergency

Filtration System filter testing from the current TS SRs to a new TS

administrative control program. The testing requirements are being

upgraded to the latest industry standards. Filter testing in

accordance with the proposed program will ensure that Title 10, Code

of Federal Regulations, Part 50 (10 CFR 50), Appendix A, General

Design Criteria (GDC) 19 and 10 CFR 100 limits are not exceeded.

c. Other Control Room and Auxiliary Electric Equipment Room

Emergency Filtration System Changes

The SBGTS and Control Room and AEER Emergency Filtration System

are designed to mitigate the radiological consequences of previously

evaluated design bases accidents. Operation and testing of these

systems will have no effect on the probability of a design bases

accident occurring.

The proposed revisions associated with this change acknowledge

that the AEERs are required to be habitable during design bases

accidents. This is consistent with the plants design bases.

d. Editorial Changes

The proposed revisions to TS 6.2.F.7 reformat the requirement to

establish consistency with the remainder of TS 6.2.F. There are no

technical changes being proposed.

Based upon the above, the proposed amendment will not increase

the probability or consequences of any accident previously

evaluated.

2) Create the possibility of a new or different kind of accident

from any accident previously evaluated because:

a. Drywell and Suppression Chamber Purge System

No new plant equipment is being installed, and use of currently

installed plant equipment is not affected by this proposed change.

The proposed revision to TS 3.6.1.8 provides additional limitations

on the opening of the drywell and suppression chamber purge system

isolation valves.

b. Ventilation Filter Testing Program

No new plant equipment is being installed, and use of currently

installed plant equipment is not affected by this proposed change.

These proposed revisions will demonstrate operability of the Control

Room and AEER Emergency Filtration System using the latest industry

standards.

c. Other Control Room and Auxiliary Electric Equipment Room

Emergency Filtration System Changes

No new plant equipment is being installed, and use of currently

installed plant equipment is not affected by this proposed change.

These proposed revisions will demonstrate habitability of the AEER

by imposing operability requirements on the AEER recirculation

filter units.

d. Editorial Changes

The proposed revisions to TS 6.2.F.7 reformat the requirement to

establish consistency with the remainder of TS 6.2.F. There are no

technical changes being proposed.

Based upon the above, the proposed change will not create the

possibility of a new or different kind of accident or transient

previously evaluated.

3) Involve a significant reduction in the margin of safety

because:

a. Drywell and Suppression Chamber Purge System

The current TS 3.6.1.8 requirements are non-conservative with

respect to the assumptions used when evaluating steam bypass of the

suppression chamber; specifically, a maximum allowable leakage area

of 0.03 square feet with the only credible leakage path was assumed

to be suppression chamber vacuum breaker valve seat leakage. This

proposed revision to TS 3.6.1.8 will make the TS requirements

consistent with those assumptions.

b. Ventilation Filter Testing Program

These proposed revisions will ensure operability of the Control

Room and Auxiliary Electric Equipment Room (AEER) Emergency

Filtration system using the latest industry standards. Filter

testing in accordance with the proposed program will ensure that GDC

19 and 10 CFR 100 limits are not exceeded.

c. Other Control Room and Auxiliary Electric Equipment Room

Emergency Filtration System Changes

These proposed revisions will ensure operably of the control

room and AEER Emergency Filtration System by demonstrating system

performance with the control room and AEER recirculation filter

units to ensure GDC 19 limits are not exceeded.

d. Editorial Changes

The proposed revisions to TS 6.2.F.7 reformat the requirement to

establish consistency with the remainder of TS 6.2.F. There are no

technical changes being proposed.

[[Page 61841]]

Based on the above, the proposed TS change does not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

requested amendments involve no significant hazards consideration.

Local Public Document Room location:Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348

Attorney for licensee: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: October 15, 1997

Description of amendment request: The proposed amendments would

revise Technical Specification Table 4.3.7.5-1, Accident Monitoring

Instrumentation Surveillance Requirements, by deleting a footnote that

provides details concerning the calibration requirements for the

drywell hydrogen concentration analyzer and monitor.

Basis for proposed no significant hazards determination: As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration, which is presented

below:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

The drywell hydrogen concentration analyzer and monitors are

required to be operable by TS 3/4.7.5, Accident Monitoring

Instrumentation. Table 4.3.7.5-1, Accident Monitoring

Instrumentation Surveillance Requirements, includes a footnote

providing unnecessary details related to the calibration of this

specific analyzer and monitors. The footnote provides information

that was determined to put the hydrogen analyzers and monitors

outside of the design basis by limiting the range of the indication

to 0% to 4% hydrogen in the drywell. The calibration method is being

corrected to provide the correct range of 0% to 10%, and requires

this note in the TS to be changed or deleted. The footnote is

proposed to be deleted from the TS, because it provides unnecessary

detail.

Deletion of the footnote will not cause an increase in the

probability of an accident, because this instrumentation is only for

accident monitoring instrumentation and thus does not affect

accident initiators or assumptions.

Deletion of the footnote will not change the consequences of an

accident previously evaluated, because this detail in the TS does

not change the requirement of performing a channel calibration at

the specified frequency. In addition, the ability to monitor

hydrogen during an accident will not be affected by deletion of the

footnote.

Therefore, this change does not involve an increase in the

probability or consequences of an accident previously evaluated.

2) Create the possibility of a new or different kind of accident

from any accident previously evaluated because:

This is monitoring instrumentation only. Deletion of the

footnote concerning specifics on how to calibrate this

instrumentation will not affect the reliability or failure modes of

the drywell hydrogen concentration analyzer and monitors. Therefore,

this change will not create the possibility of a new or different

kind of accident from any accident previously evaluated.

3) Involve a significant reduction in the margin of safety

because:

This is monitoring instrumentation only. Deletion of the

footnote concerning specifics on how to calibrate this

instrumentation will not change the requirement to perform Channel

Calibrations at the frequency specified in the TS. The details of

how to perform a Channel Calibration on the drywell hydrogen

concentration analyzer and monitors are located in plant procedures

and are in accordance with vendor recommendations. The TS

requirements for redundancy of the instrumentation and the actions

to be taken for inoperable instrumentation are also not affected by

the deletion of this footnote.

This change to the level of information regarding this

calibration is consistent with the detail for this and other

instrumentation in NUREG-1434, Revision 1, Standard Technical

Specifications, General Electric Plants, BWR/6.

Therefore, deletion of footnote * from TS Table 4.3.7.5-1 will

not involve a reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

requested amendments involve no significant hazards consideration.

Local Public Document Room location:Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348

Attorney for licensee: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Robert A. Capra

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire

Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: October 6, 1997

Description of amendment request: The proposed amendments would

delete all references to the steam line low pressure safety injection

function.

Basis for proposed no significant hazards determination: As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration, which is presented

below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated?

Answer

Probability

Accident initiators can affect the probability of a previously

evaluated accident. The addition of a new device or piece of

equipment to the plant may introduce a new accident initiator. No

new equipment is added to the plant as a result of this change. The

proposed removal of the low steam line steam pressure will involve

removing the steam line pressure safety injection function. This

results in a reduction in the likelihood of spurious safety

injections. Spurious safety injections can result in inadvertent

ECCS [emergency core cooling system] actuations. Inadvertent ECCS

Actuation is a UFSAR [updated Final Safety Analysis Report] accident

(UFSAR 15.5.1). Therefore, this change will result in a reduction in

the probability of an accident previously evaluated.

Routine plant operating practices and conditions will not be

altered by the removal of the safety injection function. Therefore,

there is no operating practice or condition change that could

increase the probability of occurrence of a previously evaluated

accident.

There is no significant increase in the probability of an

accident previously evaluated.

Consequences

Accidents previously evaluated that could be adversely affected

are the steam line break and the feedwater line break. These

accidents will result in secondary side depressurization with

pressure reaching the current actuation setpoint. The review of

these accidents found that the consequences of the previous accident

analysis acceptance criteria remain satisfied. The specifics of the

accident analysis is discussed below.

The steam line break accident was analyzed to demonstrate short

term cooling capability. A spectrum of break sizes were evaluated to

determine the limiting break size. For smaller breaks (including the

limiting break size), the safety injection actuation on low

pressurizer pressure occurs prior to low steam line pressure safety

injection. However, for larger steam line breaks the setpoint for

low steam line pressure safety injection is reached prior to low

pressurizer pressure safety injection. The larger spectrum of breaks

were analyzed without credit for the low steam line pressure safety

injection. The results of this analysis found that there would be a

slight increase in time required for safety injection to actuate.

The low pressurizer safety injection would actuate in these

accidents due to the cooldown and depressurization of the reactor

coolant system in response to the secondary side energy removal. The

Departure from

[[Page 61842]]

Nucleate Boiling Ratios (DNBRs) were analyzed with this time delay

in safety injection. The DNBRs for these cases were found to be less

limiting than those calculated for the limiting break size.

Therefore, the removal of steam line low pressure safety injection

does not adversely affect the DNBR, fuel failure or dose

consequences of the main steam line break accident. Other acceptance

criteria would not be expected to be affected by the small change in

timing of the safety injection signal.

In addition, to the Chapter 15 accident analysis, the Chapter 6

containment response to mass and energy releases was evaluated

without credit for steam line low pressure safety injection. The

evaluation demonstrated that for steam line breaks inside of

containment, the high containment pressure safety injection set

point is reached prior to the pressure associated with steam line

low pressure safety injection. Therefore the existing containment

response evaluation is not adversely affected by the removal of the

low steam pressure safety injection. This also assures that the

existing environmental qualification envelope for McGuire is not

affected by this change. For steam line breaks outside of

containment the maximum required breaksize is 1.0 ft2, which results

in transients with safety injection caused by low pressurizer

pressure prior to low steam line pressure safety injection.

The feedwater line break accidents were analyzed to demonstrate

long term core cooling capability. During a feedwater line break,

the secondary system will depressurize if the break occurs between

the main feedwater check valve and the steam generator. However,

breaks are only required to occur at the terminal ends of feedwater

piping (i.e., at the feedwater pump or at the steam generator). For

a feedwater line break at the main feedwater pump, the main feed

check valve will prevent depressurization of the steam generator.

For a feedwater line break at the steam generator, a safety

injection on high containment pressure will occur prior to safety

injection on steam pressure. Therefore, the elimination of the steam

line low pressure safety injection does not adversely impact the

feedwater line break accident.

In summary, a review was conducted of all design basis accidents

to identify those which result in a low steam pressure safety

injection. These accidents were then evaluated to verify that the

accident analysis were within acceptance criteria. This review

revealed that all accident analysis results were within current

analysis acceptance criteria.

Therefore, there is no significant increase in the consequences

of a previously evaluated accident.

Conclusion

Elimination of the low steam line pressure safety injection

results in no significant increase in the probability or

consequences of an accident previously evaluated.

(OR)

2. Create the possibility of a new or different kind of accident

from any accident previously evaluated[?]

Answer

There is no introduction of new equipment or operating practices

that could result in a new operating condition. The plant will

continue to operate in the same method with the same complement of

equipment with the exception of the actuation logic associated with

the steam line low pressure safety injection. Therefore, there is no

new operating condition that would be expected to generate a new

sequence of events which could generate a new or different accident.

There is no new equipment that could interact with other plant

structures, systems or components.

The low pressure safety injection equipment is the only plant

equipment affected by this change. There are no new equipment

failure modes which might result in a new or different accident.

Affected accidents were evaluated to validate that the accident

sequence would not deviate in a fashion which would create a new or

different accident. The analysis of the feedwater line break and

steam line break did not reveal any new or different type of

accident.

Removal of the low steam line pressure safety injection will not

create the possibility of a new or different kind of accident from

any accident previously evaluated;

(OR)

3. Involve a significant reduction in the margin of safety?

Answer

The margin of safety relevant to this change is represented by

the margin of physical protection provided by fuel cladding and the

reactor containment. Effects of this change on the safety analysis

was described under question 1 above. The results of the analysis

demonstrate that DNBR, fuel clad integrity and containment response

were not significantly affected by the removal of low steam line

pressure safety injection. Therefore, the physical protection

provide[d] by the fuel cladding and reactor containment were not

affected by this change. Accident acceptance criteria continued to

be met without credit for the safety function. The radiological

consequences of accidents was not affected by the change.

The removal of the low steam line pressure safety injection did

not significantly reduce the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, North Carolina

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-

423, Millstone Nuclear Power Station, Unit No. 3, New London

County, Connecticut

Date of amendment request: October 15, 1997

Description of amendment request: The proposed amendment would

affect nominal trip setpoints and allowable values for Reactor Trip

System (RTS) Instrumentation Trip Setpoints Table 2.2-1, and Engineered

Safety Features Actuation System (ESFAS) Instrumentation Trip Setpoints

Table 3.3-4. In addition, the proposed amendment would (1) decrease the

reactor trip setpoint for the reactor coolant pump (RCP) low shaft

speed (underspeed trip setpoint) from 95.8 percent to 92.4 percent of

rated speed, (2) make editorial changes, and (3) change the Bases to

reflect the new methodology.

Basis for proposed no significant hazards determination: As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration, which is presented

below:

NNECO has reviewed the proposed revision in accordance with

10CFR50.92 and has concluded that the revision does not involve a

significant hazards consideration (SHC). The basis for this

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

satisfied. The proposed revision does not involve [an] SHC because

the revision would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The proposed changes to Tables 2.2-1 and 3.3-4 involve changes

from a five column format to a two column format. The RTS trip

setpoints and ESFAS trip setpoints remain unchanged with the

exception of the RCP low shaft speed trip setpoint discussed below.

Detailed operability criteria will be moved to surveillance

procedures and analysis has demonstrated that an adequate margin for

normal trip setpoints exist and safety analysis limits are preserved

in all RTS/ESFAS functions.

Changing the RCP low shaft speed trip setpoint will not change

the probability of occurrence of the event. The existing accident

analysis (Millstone Unit No. 3 FSAR [final safety analysis report]

section 15.3.2) of the complete loss of forced reactor coolant flow

remains valid for the proposed change. Therefore, the change to the

RCP low shaft speed trip setpoint does not increase the probability

or consequences of any previously analyzed accident.

In addition, the proposed changes to Tables 2.2-1 and 3.3-4 do

not alter the intent or method by which the surveillances are

conducted. Therefore, the scope of evaluation performed gives

reasonable assurance that there will not be an adverse impact on the

consequences or the probability of any previously analyzed accident.

Therefore, the proposed revision does not involve a significant

increase in the probability or consequence of an accident previously

evaluated.

2. Create the possibility of a new or different kind of accident

from any accident previously evaluated.

[[Page 61843]]

The existing design basis adequately covers the plant response

with the proposed change to the RCP low shaft speed trip setpoint.

The change does not introduce new failure modes.

The proposed changes to Tables 2.2-1 and 3.3-4 do not modify the

design or operation of any plant system. The proposed changes do not

alter the intent or method by which the surveillances are conducted,

other than adjusting the allowable values to reflect historical

instrument performance data. Therefore, the proposed revision does

not create the possibility of a new or different kind of accident

from any accident previously evaluated.

3. Involve a significant reduction in a margin of safety.

The proposed changes to Tables 2.2-1 and 3.3-4 modify the

existing five column format to a two column format to show the RTS

and ESFAS nominal trip setpoints and the process rack bistable

allowable values for individual functions. Detailed operability

criteria will be moved to the surveillance procedures. With the

exception of the low shaft speed trip discussed below, the RTS and

ESFAS setpoints remain unchanged and analysis has demonstrated that

an adequate margin for normal trip setpoints exist and safety

analysis limits are preserved in all RTS/ESFAS functions.

Since the safety limits of the design are still met, the

proposed change to the RCP low shaft speed trip setpoint does not

reduce the margin of safety.

Therefore, the proposed revision does not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed revision does not involve an SHC.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

Connecticut NRC Deputy Director: Phillip F. McKee

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,

Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis

Obispo County, California

Date of amendment requests: March 10, 1997, as supplemented by

letter dated May 20, 1997

Description of amendment requests: The proposed amendments would

revise the combined Technical Specifications (TS) for the Diablo Canyon

Power Plant Unit Nos. 1 and 2 to revise TS 3/4.4.5 and 3.4.6.2,

including associated Bases 3/4.4.5 and 3/4.4.6.2, to allow the

implementation of steam generator (SG) tube alternate repair criteria

for axial indications in the Westinghouse explosive tube expansion

(WEXTEX) region below the top of the tubesheet and below the bottom of

the WEXTEX transition that may exceed the current TS depth-based

plugging limit. The allowed primary-to-secondary operational leakage

from any one SG would be reduced from 500 gpd to 150 gpd.

Basis for proposed no significant hazards determination: As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration, which is presented

below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Probability

Of the various accidents previously evaluated, the proposed

changes only affect the steam generator tube rupture (SGTR) event

evaluation and the postulated steam line break (SLB) accident

evaluation. Loss-of-coolant accident (LOCA) conditions cause a

compressive axial load to act on the tube. Therefore, since the LOCA

tends to force the tube into the tubesheet rather than pull it out,

it is not a factor in this amendment request. Another faulted load

consideration is a safe shutdown earthquake (SSE); however, the

seismic analysis of Series 51 steam generators has shown that axial

loading of the tubes is negligible during an SSE.

For the SGTR event, the required structural margins of the steam

generator tubes will be maintained by the presence of the tubesheet.

Tube rupture is precluded for cracks in the Westinghouse explosive

tube expansion (WEXTEX) region due to the constraint provided by the

tubesheet. Therefore, Regulatory Guide (RG) 1.121, ``Bases for

Plugging Degraded PWR Steam Generator Tubes,'' margins against burst

are maintained for both normal and postulated accident conditions.

The W* length supplies the necessary resistive force to preclude

pullout loads under both normal operating and accident conditions.

The contact pressure results from the WEXTEX expansion process,

thermal expansion mismatch between the tube and tubesheet and from

the differential pressure between the primary and secondary side.

Therefore, the proposed change results in no significant increase in

the probability of the occurrence of an SGTR or SLB accident. 1

The proposed changes do not affect other systems, structures,

components or operational features. Therefore, based on the above

evaluation, the proposed changes do not involve a significant

increase in the probability of an accident previously evaluated.

Consequences

The consequences of an SGTR event are affected by the primary-

to-secondary leakage flow during the event. Primary-to-secondary

leakage flow through a postulated broken tube is not affected by the

proposed change since the tubesheet enhances the tube integrity in

the region of the WEXTEX expansion by precluding tube deformation

beyond its initial expanded outside diameter. The resistance to both

tube rupture and collapse is strengthened by the tubesheet in that

region. At normal operating pressures, leakage from primary water

stress corrosion cracking (PWSCC) in the W* length is limited by

both the tube-to-tubesheet crevice and the limited crack opening

permitted by the tubesheet constraint. Consequently, negligible

normal operating leakage is expected from cracks within the

tubesheet region.

SLB leakage is limited by leakage flow restrictions resulting

from the crack and tube-to-tubesheet contact pressures that provide

a restricted leakage path above the indications and also limit the

degree of crack face opening compared to free span indications. The

total leakage, that is, the combined leakage for all such tubes,

plus the combined leakage developed by any other ARC, must be below

the maximum allowable SLB leak rate limit, such that off-site doses

are maintained less than 10 CFR 100 guideline values.

Therefore, based on the above evaluation, the proposed changes

do not involve a significant increase in the consequences of an

accident previously evaluated.

2. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed changes do not introduce any changes or mechanisms

that create the possibility of a new or different kind of accident.

Tube bundle integrity is expected to be maintained for all plant

conditions upon implementation of the proposed steam generator

alternate tube plugging criteria.

WCAP-14797, Revision 1, ``Generic W* Tube Plugging Criteria for

51 Series Steam Generator Tubesheet Region WEXTEX Expansions,''

requires that any tubes with indications identified using the bobbin

coil probe during the bobbin sampling plan also be inspected with

the RPC coil throughout the W* length of the tubes. The use of the

RPC will: (a) identify any new or non-expected degradation mode that

may not be identified using the bobbin coil probe, and (b) confirm

and characterize the bobbin coil indication.

These changes do not introduce any new equipment or any change

to existing equipment. No new effects on existing equipment are

created nor are any new malfunctions introduced.

Therefore, based on the above evaluation, the proposed changes

do not create the possibility of a new or different kind of accident

from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes maintain the required structural margins of

the steam

[[Page 61844]]

generator tubes for both normal and accident conditions. RG 1.121 is

used as the basis in the development of the W* alternate tube

plugging criteria for determining that steam generator tube

integrity considerations are maintained within acceptable limits. RG

1.121 describes a method acceptable to the NRC staff for meeting

General Design Criteria 14, 15, 31, and 32 by reducing the

probability and consequences of an SGTR. RG 1.121 concludes that by

determining the limiting safe conditions of tube wall degradation

beyond which tubes with unacceptable cracking, as established by

inservice inspection, should be removed from service or repaired,

the probability and consequences of an SGTR are reduced. This RG

uses safety factors on loads for tube burst that are consistent with

the requirements of Section III of the ASME Code.

For primarily axially oriented cracking located within the

tubesheet, tube burst is precluded due to the presence of the

tubesheet. WCAP-14797 defines a length, W*, of degradation free

expanded tubing that provides the necessary resistance to tube

pullout due to the pressure induced forces (with applicable safety

factors applied). Application of the W* criteria will preclude

unacceptable primary-to-secondary leakage during all plant

conditions. The methodology for determining leakage provides for

large margins between calculated and actual leakage values in the W*

criteria.

Plugging of the steam generator tubes reduces the reactor

coolant flow margin for core cooling. Implementation of the proposed

changes are expected to result in plugging of fewer tubes than with

the current criteria. Thus, implementation of the proposed changes

will maintain the margin of flow that may have otherwise been

reduced by tube plugging.

Based on the above, it is concluded that the proposed changes do

not result in a significant reduction of margin with respect to

plant safety as defined in the FSAR Update or bases of the plant

Technical Specifications.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment requests involve no significant hazards consideration.

Local Public Document Room location: California Polytechnic State

University, Robert E. Kennedy Library, Government Documents and Maps

Department, San Luis Obispo, California 93407

Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and

Electric Company, P.O. Box 7442, San Francisco, California 94120

NRC Project Director: William H. Bateman

Philadelphia Electric Company, Docket No. 50-352, Limerick

Generating Station, Unit 1, Montgomery County, Pennsylvania

Date of amendment request: October 24, 1997

Description of amendment request: The proposed Technical

Specifications (TS) changes would revise TS Section 3/4.1.3.6 to exempt

control rod 50-27 from the coupling test for the remainder of Cycle 7.

Basis for proposed no significant hazards determination: As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration, which is presented

below:

1. The proposed Technical Specifications (TS) change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated. The probability of occurrence of

the analyzed Control Rod Drop Accident (CRDA) is not increased by

operating with the subject control blade in a condition not known to

be coupled since the compensatory measures will assure that the

blade will remain fully inserted below 10% rated thermal power where

the CRDA is a concern. Monitoring of nuclear instrumentation

responses in the vicinity of the blade when the drive is withdrawn

above 10% power will assure the blade is tracking with the drive

with no potential to stick and then drop. Scram impact forces from

an uncoupled control rod are of insufficient energy to dislodge the

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

integrity. No reduction of system or equipment redundancy is

involved.

The CRDA analyzed in the Safety Analysis Report (SAR) remains

the limiting rod drop accident, and its consequences are unaffected

by operation of the subject blade in the proposed manner. Operation

of the control blade as described, i.e., withdrawn no further than

the 46 position and in a condition not known to be coupled, has no

adverse effect on scram performance in response to any other

postulated accident. The scram insert motion of the rod is not

affected by the potentially uncoupled condition, and since the rod

is already partially inserted at position 46, it should have a

slightly better negative reactivity insertion characteristic.

Therefore, no potential to increase onsite or offsite radiological

consequences beyond those previously analyzed in the SAR is created.

Operating the subject control blade in a condition not known to

be coupled does not result in any onsite or offsite radiological

consequences different from those previously analyzed in the SAR.

The subject control blade will be fully inserted below 10% thermal

power where the CRDA is a concern and will be monitored during drive

withdrawal above 10% thermal power to assure it is tracking with the

drive. Scram performance is not adversely affected by operation from

the near full-out position of 46. Hence, no new failure modes are

created and consequences of any postulated failures are not

increased.

Therefore, the proposed TS change does not involve an increase

in the probability or consequences of an accident previously

evaluated.

2. The proposed TS change does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The Safety Analysis Report (SAR) analyzed Control Rod Drop

Accident (CRDA) remains the only type of accident initiated (or

contributed to) by the control rod drive/control blade interface.

The compensatory actions to be taken when operating the subject

blade in a condition not verified to be coupled assure that no new

types of accidents can occur. The subject control blade will be

fully inserted below 10% thermal power where the CRDA is a concern

and will be monitored during drive withdrawal above 10% thermal

power to assure it is tracking with the drive. Scram performance is

not adversely affected by operation from the near full-out position

of 46. Since no adverse effect on insertion or scram performance is

expected, the previously analyzed accidents encompass any potential

consequence of operating with an uncoupled control blade.

The compensatory actions to be taken when operating the subject

blade in a condition not verified to be coupled assure that no new

failure modes are created, and, therefore, no new type of equipment

malfunction is introduced by operating the subject control blade in

the proposed manner.

Therefore, the proposed TS change does not create the

possibility of a new or different kind of accident from any accident

previously evaluated.

3. The proposed TS change does not involve a significant

reduction in a margin of safety.

Operation with the subject control blade in a condition not

known to be coupled for the remainder of Cycle 7 at LGS [Limerick

Generating Station] Unit 1, but with the compensatory actions

described below, does not reduce the existing margin of safety

determined by the analysis of the Control Rod Drop Accident (CRDA).

The CRDA analyzed in the Safety Analysis Report (SAR) remains

bounding in that the subject rod will be fully inserted below 10%

rated thermal power where the CRDA is a concern. Above 10% power,

when the associated drive is withdrawn, the nuclear instrumentation

in the vicinity of the blade will be monitored to assure the blade

tracks with the drive, providing assurance that the position of the

blade can be ascertained by the drive position. If the control blade

can not be verified to have followed the drive, then the rod shall

be completely inserted and the control rod directional valves

disarmed in accordance with existing TS requirements. To minimize

any scram impact loadings, the blade will be operated at the near

full-out position of 46 except for intermediate positions

temporarily occupied during standard rod withdrawal sequences.

Operating the subject control blade in the proposed manner will have

no adverse effect on insertion or scram performance of the blade and

will preserve the margin of safety.

[[Page 61845]]

Therefore, the proposed TS change does not involve a reduction

in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location:Pottstown Public Library, 500

High Street, Pottstown, PA 19464

Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, PA 19101

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-272, Salem

Nuclear Generating Station, Unit No. 1, Salem County, New Jersey

Date of amendment request: October 6, 1997

Description of amendment request: The amendment to the Technical

Specifications would increase the allowable band for control and

shutdown rod demanded position versus indicated position from plus or

minus 12 steps to plus or minus 18 steps when the power level is not

greater than 85% rated thermal power. The amendment is identical to

Amendment 183 for Salem Unit 2, which was issued September 10, 1997, as

an exigent amendment.

Basis for proposed no significant hazards determination: As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration, which is presented

below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change to the rod misalignment criteria of [plus or

minus] 18 steps for core powers equal to or below 85% of RATED

THERMAL POWER (RTP) does not increase the probability of previously

evaluated accidents. Increasing the magnitude of the allowed control

rod misalignment is not a contributor to the mechanistic cause of an

accident evaluated in any accident analysis. The magnitude of

control rod indicated misalignment is a parameter used to establish

the initial conditions for accident evaluation.

The proposed increase in the allowable rod misalignment from the

current [plus or minus] 12 steps for reactor powers equal to or less

than 85% RTP does not involve a significant increase in the

consequence of any previously evaluated accident. Rod misalignment

affects power distribution, shutdown margin and the ejected rod

accident. An extension of the allowable rod misalignment above and

below 85% RTP has been analyzed in Westinghouse WCAP-14672. As

provided in WCAP-14672, above 85% the allowable misalignment is

governed by the available peaking factor margins as determined by

flux maps.

[Public Service Electric & Gas] PSE&G is simplifying the

proposed change by keeping the currently allowed [plus or minus] 12

step misalignment in Technical Specifications 3.1.3.1 and 3.1.3.2.1

for reactor power greater than 85% RTP.

The PSE&G proposed change is to allow [plus or minus] 18 steps

misalignments in Technical Specifications 3.1.3.1 and 3.1.3.2.1 for

reactor power less than or equal to 85% RTP. As demonstrated in

WCAP-14672, for reactor powers less than 85% RTP, the available

peaking factor margin increases faster than any penalty associated

with a [plus or minus] 18 step misalignment.

As described in Section 4.0 of the Westinghouse WCAP, a

conservative penalty factor has been applied to the rod insertion

allowance (RIA) of the shutdown margin calculation to account for

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

of [plus or minus] 18 steps). This conservative penalty factor is

applied as part of the reload analysis in order to satisfy Technical

Specification 3.1.1.1.

In addition to the normal, or Condition 1, operational

transients, the impacts of increased rod misalignment on Condition

II, III and IV accident analysis have also been evaluated. The

proposed increase in rod misalignment does not have a significant

effect on any moderator or Doppler reactivity coefficients or

defects, boron worth or reactor kinetics parameters.

To account for the potential increase in ejected rod parameters,

conservative penalty factors have been applied to the reload safety

evaluation to cover the additional [plus or minus] 6 step

misalignment. Margin is available in the reload safety analysis to

accommodate this impact.

Therefore, the proposed amendment does not increase the

probability or consequences of any accident previously evaluated.

2. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

0No new accident scenarios, failure mechanisms or limiting

single failures are introduced as a result of the proposed change to

the rod misalignment criteria of [plus or minus] 18 steps below 85%

RTP. The implementation of the proposed rod misalignment criteria

will have no adverse effect on the performance of any other safety

related system. Therefore, the proposed amendment does not create

the possibility of a new or different kind of accident from any

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in the margin of

safety. The Technical Specifications allowed increase in peaking

factors as power is reduced accommodates the peaking factor penalty

associated with the additional [plus or minus] 6 step misalignment

for core powers equal to or less than 85% RTP. Therefore, there is

no change to the peaking factors assumed in the safety analysis. In

addition to peaking factors, there is no change in any other current

limit input into the safety analysis. As the input, or initial

conditions, of the safety analysis have not changed, there is no

reduction in the margin to safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079

Public Service Electric & Gas Company, Docket No. 50-272, Salem

Nuclear Generating Station, Unit No. 1, Salem County, New Jersey

Date of amendment request: October 14, 1997

Description of amendment request: The proposed amendment will

modify the Salem Unit 1 Technical Specification (TS) 3.4.6.3, ``Primary

Coolant System Pressure Isolation Valves Limiting Condition for

Operation,'' to be consistent with Salem Unit 2 TSs.

Basis for proposed no significant hazards determination: As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration, which is presented

below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The majority of the proposed changes, as described above, are

editorial in nature. Rewording, and reformatting the Limiting

Condition for Operation, including the surveillance requirements do

not involve a significant increase to the probability or

consequences of an accident.

Those substantive changes involving the addition of (1) new

reactor coolant system pressure isolation valves, (2) providing for

a shorter test frequency upon entry into Mode 4, and (3) adding a

new surveillance test requirement, do not increase the probability

or consequences of an accident. These changes ensure that the system

and components needed to prevent and minimize the effects of inter-

system loss of coolant are properly identified in the Technical

Specifications.

Although pressure isolation valves are being added to the

Technical Specification table, these valves were already included in

the IST [inservice testing] program as pressure isolation valves and

were being tested as such. The proposed changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

[[Page 61846]]

2. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed change, as described above, does not physically

alter the facility or the operation of the facility. The majority of

the changes are editorial in nature and provide for improvement in

the human factors of the Technical Specifications, while properly

identifying all the pressure isolation valves in the Technical

Specifications. The addition of valves into the Technical

Specification is an administrative change that improves the quality

of the LCO [limiting condition for operation], but does not add

components to the facility.

The proposed changes do not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The margin of safety, as defined in the bases for any technical

specifications, depend upon proper identification of equipment and

performance of the proper surveillance requirements to demonstrate

equipment operability. The proposed change will ensure that the

proper valves are identified and tested in accordance with the

Technical Specification requirements.

The proposed changes do not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business

Unit - N21, P.O. Box 236, Hancocks Bridge, NJ 08038

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-

311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of amendment request: October 21, 1997

Description of amendment request: The proposed amendment revises

Technical Specification Tables 3.3-1 and 4.3-1 to require that

Functional Unit, 2. Power Range, Neutron Flux,'' be operable

in Mode 3, as well as in Modes 1 and 2. The change is being proposed

because the licensee has determined that the power range nuclear

instrumentation should be operable in Mode 3 whenever the reactor trip

system breakers are in the closed position and the control rods are

capable of being withdrawn.

Basis for proposed no significant hazards determination: As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration, which is presented

below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The requirement for operability of a trip and the surveillance

requirements to ensure the functionality of the trip are independent

of the probability of an accident previously evaluated. The accident

that this trip is intended to mitigate is the Rod Withdrawal from

Subcriticality event. The surveillance procedure and the requirement

for the trip to be operational when the Control Rod Drive System is

capable of rod movement mitigate the consequences of this event, and

do not increase the probability of a rod withdrawal from

subcritical.

Therefore, the probability and consequences of an accident

previously evaluated are not significantly increased.

2. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed changes do not involve any modifications to

existing plant equipment, do not alter the function of any plant

systems, do not introduce any new operating configurations or new

modes of plant operation, or change the safety analyses. The

proposed change is intended to ensure that the trip function is

available and will perform as designed in the event of a previously

evaluated event.

The proposed changes will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change does not reduce the margin of safety,

because assurance of the operability of the trip function is

increased by the proposed change.

Based on the above, PSE&G [Public Service Electric & Gas

Company] has determined that the proposed changes do not involve a

significant hazards consideration.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Salem Free Public library, 112

West Broadway, Salem, NJ

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business

Unit - N21, P.O. Box 236, Hancocks Bridge, NJ 08038

NRC Project Director: John F. Stolz

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania

Power Company, Toledo Edison Company, Docket No. 50-440, Perry

Nuclear Power Plant, Unit 1, Lake County, Ohio

Date of amendment request: August 28, 1997

Description of amendment request: The proposed change would revise

Technical Specification 3.4.11, ``Reactor Coolant System (RCS) Pressure

and Temperature (P/T) Limits,'' to incorporate the new P/T curves,

which were provided by General Electric Nuclear Energy in report number

GE-NE-B1301793-01, ``Perry Unit 1 RPV Surveillance Materials Testing

and Analysis.''

Basis for proposed no significant hazards determination: As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration which is presented

below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change will provide for approved P/T limit curves

which are valid through 9 effective full-power years (EFPY) and 18

EFPY. This change will not affect any Safety Limits, Power

Distribution Limits, or Limiting Conditions for Operation. The

proposed changes incorporate operating limits which provide margin

to brittle failure of the reactor vessel based on testing of the

irradiated reactor vessel materials (base metal, weld material, and

heat affected zone material). The limits ensure that adequate safety

margins against nonductile or rapidly propagating failure exist

during normal operation, anticipated operational occurrences, and

system hydrostatic tests. The specimens have been tested and

analyzed in accordance with 10 CFR 50, Appendices G and H, using the

methods described in Generic Letter 88-11 and Regulatory Guide 1.99

Revision 2. The predicted lowest upper shelf energy at 32 EFPY was

greater than the minimum required by 10 CFR 50, Appendix G. The

adjusted reference temperature for the limiting material was lower

than the 200 degree Fahrenheit limit required by Regulatory Guide

1.99 Revision 2. As such, the integrity of the reactor pressure

coolant boundary is maintained. The changes will result in

equivalent or more conservative limits on reactor vessel pressure as

a function of temperature for all operational conditions

(hydrostatic and leak testing, non-nuclear heatup/cooldown, and core

critical operations). The methodology used to derive these values

produces limits which continue to ensure that sufficient margin is

maintained to meet the criteria of GDC 31, ``Fracture Prevention of

Reactor Coolant Pressure Boundary.'' There are no plant

modifications associated with this change and no new or revised

system interfaces. The proposed

[[Page 61847]]

changes do not increase the probability of occurrence or

consequences previously evaluated because the temperature shifts are

well within equipment operating ranges. As such, there is no

increase in the probability of occurrence or the consequences of an

accident previously evaluated.

2. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed changes do not create the possibility of a new or

different kind of accident from any accident previously evaluated

because the proposed changes do not involve any new modes of

operation. The only change will be operation of the plant within

operating pressure limits which are determined in a more

conservative manner. Therefore, no new failure mode or accident

sequence is introduced by this change.

The testing and analysis meets 10 CFR 50, Appendices G and H,

requirements; therefore, no new accident types, such as brittle

fracture of a reactor pressure coolant boundary component is

postulated. The adjusted reference temperature and upper shelf

energy predicted at 32 EFPY are well within the limits of 10 CFR 50,

Appendices G and H. Therefore, the possibility of an accident of a

new or different type than any previously evaluated is not created.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The P/T limits are established to provide acceptable margins for

the operation of the reactor coolant system during heat up and cool

down, criticality, and hydrotest conditions. Technical Specification

3.4.11 limits the rates of change of temperature and pressure to

values consistent with the fracture toughness requirements of 10 CFR

50, Appendices G and H, and ASME Boiler and Pressure Vessel Code

Section III Appendix G. The bases section for Technical

Specification 3.4.11 refers to 10 CFR 50, Appendices G and H, and

ASME Code Section III Appendix G. Changes in these limits are

necessary because the fracture toughness properties of ferritic

materials in the reactor vessel change as a function of reactor

operating time. The specific requirements for fracture toughness and

reactor vessel material surveillance that must be considered in

developing the P/T limits are defined by 10 CFR 50, Appendices G and

H. The specific limits defined by 10 CFR 50, Appendices G and H, set

the margin of safety for the reactor pressure vessel coolant

boundary. Since the testing and analysis of the vessel specimens

meet the requirements and limits defined in 10 CFR 50, Appendices G

and H, the margin of safety as defined in the basis for Technical

Specification 3.4.11 is not reduced. The revised curves are based on

the latest NRC guidelines along with actual neutron fluence data for

Perry. The new limits conservatively account for irradiation

embrittlement effects, thereby maintaining the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW., Washington, DC 20037

NRC Project Director: Gail H. Marcus

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania

Power Company, Toledo Edison Company, Docket No. 50-440, Perry

Nuclear Power Plant, Unit 1, Lake County, Ohio

Date of amendment request: September 8, 1997

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 5.2.2.e, ``Organization - Unit

Staff,'' by removing the reference to the NRC Policy Statement on

working hours. Administrative procedures will be developed to limit the

working hours of unit staff who perform safety-related functions.

Basis for proposed no significant hazards determination: As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration which is presented

below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change to TS 5.2.2.e only alters the administrative

location of and the regulatory controls applicable to unit staff

specific overtime limits and working hours. Overtime will remain

controlled by plant administrative procedures. Changes to the

relocated overtime limits and working hours will be subject to

review and evaluation under 10 CFR 50.59, ``Changes, Tests and

Experiments.'' There is not an increase in the probability of an

accident previously evaluated because no change is being made to any

accident initiator. No previously analyzed accident scenario is

changed, and initiating conditions and assumptions remain as

previously analyzed.

There is not an increase in the radiological consequences of an

accident previously evaluated because the proposed change does not

affect accident conditions or assumptions used in evaluating the

radiological consequences of an accident. The proposed change does

not alter the source term, containment isolation, or allowable

radiological releases. Therefore, there is no increase in the

radiological consequences of an accident previously evaluated.

2. The proposed change does not create the possibility or a new

or different kind of accident from any accident previously

evaluated.

The proposed change does not create the possibility of a new or

different kind of accident from any accident previously evaluated

because the proposed change does not change the way the plant is

operated, and no new or different failure modes have been defined

for any plant system or component important to safety, nor has any

limiting single failure been identified as a result of the proposed

change. No new or different types of failures or accident initiators

are introduced by the proposed change.

The proposed change to TS 5.2.2.e only alters the administrative

location of and the regulatory controls applicable to unit staff

specific overtime limits and working hours. Therefore, there is no

possibility created for a new or different kind of accident.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change does not involve a reduction in a margin of

safety because unit staff overtime is not an input in the

calculation of a safety margin with regard to Technical

Specification Safety Limits, Limiting Safety System Settings, other

Technical Specification Limiting Conditions for Operation, the

Operational Requirements Manual, or other previously defined margins

for any structure, system, or component important to safety. The

proposed change to TS 5.2.2.e only alters the administrative

location of and the regulatory controls applicable to unit staff

specific overtime limits and working hours.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW., Washington, DC 20037

NRC Project Director: Gail H. Marcus

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak

Steam Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: October 24, 1997

Brief description of amendments: Change to the core safety limit

curves and overtemperature N-16 reactor trip function setpoints to

support operation with Unit 1, cycle 7 core configuration.

Basis for proposed no significant hazards determination: As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration, which is presented

below:

1. Do the proposed changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

A. Revision to the Unit 1 Core Safety Limits

[[Page 61848]]

Analyses of reactor core safety limits are required as part of

reload calculations for each cycle. TU Electric has performed the

analyses of the Unit 1, Cycle 7 core configuration to determine the

reactor core safety limits. The methodologies and safety analysis

values result in new operating curves which, in general, permit

plant operation over a similar range of acceptable conditions. This

change means that if a transient were to occur with the plant

operating at the limits of the new curve, a different temperature

and power level might be attained than if the plant were operating

within the bounds of the old curves.

However, since the new curves were developed using NRC approved

methodologies which are wholly consistent with and do not represent

a change in the Technical Specification BASES for safety limits, all

applicable postulated transients will continue to be properly

mitigated. As a result, there will be no significant increase in the

consequences, as determined by accident analyses, of any accident

previously evaluated.

B. Revision to Unit 1 Overtemperature N-16 Reactor Trip

Setpoints

As a result of changes discussed, the Overtemperature reactor

trip setpoint has been recalculated. These trip setpoints help

ensure that the core safety limits are protected and that all

applicable limits of the safety analysis are met.

Based on the calculations performed, no significant changes to

the safety analysis values for Overtemperature reactor trip setpoint

were required. The f(deltaI) trip reset function was revised due to

more top-skewed axial power distributions predicted for this cycle.

The analyses performed show that, using the TU Electric

methodologies, all applicable limits of the safety analysis are met.

This setpoint provides a trip function which allows the mitigation

of postulated accidents and has no impact on accident initiation.

Therefore, the changes in safety analysis values do not involve an

increase in the probability of an accident and, based on satisfying

all applicable safety analysis limits, there is no significant

increase in the consequences of any accident previously evaluated.

In addition, sufficient operating margin has been maintained in

the overtemperature setpoint such that the risk of turbine runbacks

or reactor trips due to upper plenum flow anomalies or other

operational transients will be minimized, thereby, reducing

potential challenges to the plant safety systems.

SUMMARY

The changes in the amendment request applies NRC approved

methodologies to changes in safety analysis values, new core safety

limits and new N-16 setpoint and parameter values to assure that all

applicable safety analysis limits have been met. The potential for

an operational transient to occur has not been affected and there

has been no significant impact on the consequences of any accident

previously evaluated.

2. Do the proposed changes create the possibility of a new or

different kind of accident from any accident previously evaluated?

The proposed changes involve the calculation of new reactor core

safety limits and overtemperature reactor trip setpoint resets. As

such, the changes play an important role in the analysis of

postulated accidents but none of the changes effect plant hardware

or the operation of plant systems in a way that could initiate an

accident. Therefore, the proposed changes do not create the

possibility of a new or different kind of accident from any accident

previously evaluated.

3. Do the proposed changes involve a significant reduction in a

margin of safety?

In reviewing and approving the methods used for safety analyses

and calculations, the NRC has approved the safety analysis limits

which establish the margin of safety to be maintained. While the

actual impact on safety is discussed in response to question 1, the

impact on margin of safety is discussed below:

A. Revision to the Unit 1 Reactor Core Safety Limits

The NRC-approved TU Electric reload analysis methods have been

used to determine new reactor core safety limits. All applicable

safety analysis limits have been met. The methods used are wholly

consistent with Technical Specification BASES 2.1 which is the bases

for the safety limits. In particular, the curves assure that for

Unit 1, Cycle 7, the calculated DNBR is no less than the safety

analysis limit and the average enthalpy at the vessel exit is less

than the enthalpy of saturated liquid. The acceptance criteria

remains valid and continues to be satisfied; therefore, no change in

a margin of safety occurs.

B. Revision to Unit 1 Overtemperature N-16 Reactor Trip

Setpoints

Because the reactor core safety limits for CPSES Unit 1, Cycle 7

are recalculated, the Reactor Trip System instrumentation setpoint

values for the Overtemperature N-16 reactor trip setpoint which

protect the reactor core safety limits must also be recalculated.

The Overtemperature N-16 reactor trip setpoint helps prevent the

core and Reactor Coolant System from exceeding their safety limits

during normal operation and design basis anticipated operational

occurrences. The most relevant design basis analysis in Chapter 15

of the CPSES Final Safety Analysis Report (FSAR) which is affected

by the Overtemperature reactor trip setpoint is the Uncontrolled Rod

Cluster Control Assembly Bank Withdrawal at Power (FSAR Section

15.4.2). This event has been analyzed with the new safety analysis

value for the Overtemperature reactor trip setpoint to demonstrate

compliance with event specific acceptance criteria. Because all

event acceptance criteria are satisfied, there is no degradation in

a margin of safety.

The nominal Reactor Trip System instrumentation setpoints values

for the Overtemperature N-16 reactor trip setpoint (Technical

Specification Table 2.2-1) are determined based on a statistical

combination of all of the uncertainties in the channels to arrive at

a total uncertainty. The total uncertainty plus additional margin is

applied in a conservative direction to the safety analysis trip

setpoint value to arrive at the nominal and allowable values

presented in Technical Specification Table 2.2-1. Meeting the

requirements of Technical Specification Table 2.2-1 assures that the

Overtemperature reactor trip setpoint assumed in the safety analyses

remains valid. The CPSES Unit 1, Cycle 7 Overtemperature reactor

trip setpoint is not significantly different from the previous

cycle, and thus provides operational flexibility to withstand mild

transients without initiating automatic protective actions. Although

the value of the f(deltaI) trip reset function setpoint is

different, the Reactor Trip System instrumentation setpoint values

for the Overtemperature N-16 reactor trip setpoint are consistent

with the safety analysis assumptions which have been analytically

demonstrated to be adequate to meet the applicable event acceptance

criteria. Thus, there is no reduction in a margin of safety.

Using the NRC approved TU Electric methods, the reactor core

safety limits are determined such that all applicable limits of the

safety analyses are met. Because the applicable event acceptance

criteria continue to be met, there is no significant reduction in

the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location:University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, TX 76019

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and

Bockius, 1800 M Street, N.W., Washington, DC 20036

NRC Project Director: James W. Clifford, Acting

Previously Published Notices Of Consideration Of Issuance Of

Amendments To Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, And Opportunity For A Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

[[Page 61849]]

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire

Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: October 13, 1997

Description of amendment request: The proposed amendments would

support replacement of the three safety-related wide range level

instruments. The engineered safety features trip setpoint for the

refueling water automatic switchover to recirculation would be revised

to account for the difference in instrument uncertainty associated with

wide range level instruments and provide additional response time

margin.

Date of publication of individual notice in Federal Register:

October 22, 1997 (62 FR 54859)

Expiration date of individual notice: November 21, 1997

Local Public Document Room location: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, North Carolina

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287,

Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South

Carolina

Date of amendment request: October 20, 1997

Description of amendment request: The proposed amendments would

allow use of a rerolling process as an additional repair method for

tube degradation found in the tubesheet region. The rerolling method is

designed to ensure that the area of degradation will not serve as a

pressure boundary once the repair roll is installed, thus permitting

the tube to remain in service.

Date of publication of individual notice in Federal Register:

October 28, 1997 (62 FR 55835)

Expiration date of individual notice: Comment period ends November

12, 1997; Notice period ends November 28, 1997

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina

Notice Of Issuance Of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

with the standards and requirements of the Atomic Energy Act of 1954,

as amended (the Act), and the Commission's rules and regulations. The

Commission has made appropriate findings as required by the Act and the

Commission's rules and regulations in 10 CFR Chapter I, which are set

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

10 CFR 51.12(b) and has made a determination based on that assessment,

it is so indicated.

For further details with respect to the action see (1) the

applications for amendment, (2) the amendment, and (3) the Commission's

related letter, Safety Evaluation and/or Environmental Assessment as

indicated. All of these items are available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the local public document rooms for

the particular facilities involved.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,

Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,

Pennsylvania

Date of application for amendments: October 4, 1997

Brief description of amendments: These amendments revise the

surveillance requirements in Technical Specifications (TSs) 4.1.2.3.1,

4.1.2.4.1, 4.5.2.b, and 4.6.2.1.b and associated Bases. The subject

surveillance requirements are applicable to the charging/high-head

safety injection pumps, low-head safety injection pumps, and the

containment quench spray pumps. The proposed changes replace the

current specific test acceptance criteria contained in these

surveillance requirements with requirements to verify pump performance

in accordance with the inservice testing program, the emergency core

cooling system flow analysis, or the containment integrity safety

analysis, as applicable. The proposed changes also make minor editorial

changes in these TSs and make conforming changes in the TS Index pages.

Date of issuance: October 28, 1997

Effective date: Both units, as of date of issuance, to be

implemented within 60 days.

Amendment Nos.: 207, 86

Facility Operating License Nos. DPR-66 and NPF-73: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 18, 1996 (61

FR 66706) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 28, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA 15001

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Mississippi Power &

Light Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit

1, Claiborne County, Mississippi

Date of application for amendment: May 27, 1997, supplemented by

October 6, 1997

Brief description of amendment: The amendment eliminated selected

response time testing (RTT) surveillance requirements (SRs) from the

Technical Specifications (TSs) for certain components of the following

systems: reactor protection system (SR 3.3.1.1.15), primary containment

and drywell isolation instrumentation (SR 3.3.6.1.8), and emergency

core cooling system (SRs 3.5.1.8 and 3.5.2.7).

Date of issuance: November 5, 1997

Effective date: November 5, 1997

Amendment No.: 133

Facility Operating License No. NPF-29: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: June 18, 1997 (62 FR

33122) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 5, 1997.No significant

hazards consideration comments received: No.

Local Public Document Room location: Judge George W. Armstrong

Library, 220 S. Commerce Street, Natchez, MS 39120

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam

Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: July 17, 1996, as supplemented October

14, 1997

Brief description of amendment: The amendment revises

[[Page 61850]]

Facility Operating License No. NPF-38 to reflect the name change

from Louisiana Power & Light Company to Entergy Louisiana, Inc.

Date of issuance: November 3, 1997

Effective date: November 3, 1997, to be implemented within 60 days.

Amendment No.: 134

Facility Operating License No. NPF-38: Amendment revised

Facility Operating License No. NPF-38.

Date of initial notice in Federal Register: April 9, 1997 (62 FR

29749) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 3, 1997. The letter

dated October 14, 1997, provided clarifying information which did not

alter the initial no significant hazards determination. No significant

hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, LA 70122

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: March 27, as supplemented April

3, May 1, and August 20, 1997.

Brief description of amendment: Change Technical Specifications

(TS) to permanently establish a primary-to-secondary leak rate of 150

gallons per day through any one steam generator and specify the steam

generator tube inservice inspection requirements for pit-like

intergranular attack degradation in the ``B'' Once-Through-Steam-

Generator.

Date of issuance: October 28, 1997

Effective date: October 28, 1997

Amendment No.: 158

Facility Operating License No. DPR-72: Amendment revised the TS.

Date of initial notice in Federal Register: June 4, 1997 (62 FR

30632) The August 20, 1997, letter provided clarifying information that

did not affect the initial no significant hazards consideration. The

Commission's related evaluation of the amendment is contained in a

Safety Evaluation dated October 28, 1997. No significant hazards

consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629

Florida Power and Light Company, Docket Nos. 50-250 and 50-251,

Turkey Point Plant Units 3 and 4, Dade County, Florida

Date of application for amendments: August 27, 1997

Brief description of amendments: The admendments change the

Administrative Section of the Technical Specifications (TS) to allow

the use of 12-hour shifts.

Date of issuance: October 27, 1997

Effective date: October 27, 1997

Amendment Nos: 194 and 188Facility Operating Licenses Nos. DPR-31

and DPR-41: Amendments revised the TS.

Date of initial notice in Federal Register: September 24, 1997 (62

FR 50006) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 27, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,

Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County,

Michigan

Date of application for amendments: March 26, 1997.

Brief description of amendments: The amendments modify surveillance

4.7.5.1.e.2 which requires verification of the control room ventilation

system autostart function.

Date of issuance: October 28, 1997

Effective date: October 28, 1997, with full implementation within

45 days.

Amendment Nos.: 218 and 202

Facility Operating License Nos. DPR-58 and DPR-74: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: May 21, 1997 (62 FR

27796) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 28, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

Library, 500 Market Street, St. Joseph, Michigan 49085

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone

Nuclear Power Station, Unit 1, New London County, Connecticut

Date of application for amendment: May 15, 1997

Brief description of amendment: The amendment revises Technical

Specification Sections 3.1 and 4.1, ``Reactor Protection System,'' and

the associated Bases to remove run mode intermediate range monitor high

flux/inoperative with the associated average power range monitor

downscale scram trip function. The amendment also makes other editorial

revisions.

Date of issuance: October 27, 1997

Effective date: As of the date of issuance, to be implemented

within 60 days.

Amendment No.: 103

Facility Operating License No. DPR-21: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 18, 1997 (62 FR

33127) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 27, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut 06360 and at the Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

Northeast Nuclear Energy Company, et al., Docket No. 50-336,

Millstone Nuclear Power Station, Unit No. 2, New London County,

Connecticut

Date of application for amendment: September 2, 1997

Brief description of amendment: The amendment changes the Technical

Specifications (TSs) by modifying the maximum allowed primary

containment internal pressure during normal operation from 2.1 pounds

per square inch gauge (psig) to 1.0 psig. The TS Bases, Section 3/

4.6.1.4, is also updated to reflect the new maximum allowed primary

containment internal pressure during normal operation.

Date of issuance: October 27, 1997

Effective date: As of the date of issuance, to be implemented

within 30 days.

Amendment No.: 209

Facility Operating License No. DPR-65: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 24, 1997 (62

FR 50007) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 27, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut, and Waterford Library, ATTN: Vince Juliano, 49

Rope Ferry Road, Waterford, Connecticut

[[Page 61851]]

Northern States Power Company, Docket No. 50-263, Monticello

Nuclear Generating Plant, Wright County, Minnesota

Date of application for amendment: November 25, 1996, as

supplemented December 12, 1996, April 23, May 8, July 1, August 21, and

September 29, 1997

Brief description of amendment: The amendment modifies the

Technical Specification requirements associated with the Minimum

Critical Power Ratio (MCPR) safety limits for Cycle 18 based on the

cycle-specific analysis of the current mixed core of GE11/GE10 fuel

parameters.

Date of issuance: October 29, 1997

Effective date: October 29, 1997

Amendment No.: 99

Facility Operating License No. DPR-22: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 9, 1997 (62 FR

17238) The December 12, 1996, letter provided an affidavit for the

original application dated November 25, 1996. The April 23, May 8,

August 21, and September 29, 1997, letters provided clarifying

information in response to the staff's request for additional

information during a teleconference on March 18, 1997. The July 1,

1997, letter provided a nonproprietary version of the April 23, 1997,

submittal. This information was within the scope of the original

application and did not change the staff's initial proposed no

significant hazards considerations determination. Therefore, renoticing

was not warranted. The Commission's related evaluation of the amendment

is contained in a Safety Evaluation dated October 29, 1997. No

significant hazards consideration comments received: No.

Local Public Document Room location:Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

Northern States Power Company, Docket Nos. 50-282 and 50-306,

Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue

County, Minnesota

Date of application for amendments: November 27, 1996, as

supplemented August 15, September 2, and October 3, 1997

Brief description of amendments: The amendments incorporate

Combustion Engineering steam generator tube sleeve designs and

installation and examination techniques into the plant Technical

Specifications (TS). Specifically, the amendments make changes to TS

4.12, ``Steam Generator Tube Surveillance,'' and its associated Bases

Section B.4.12, ``Steam Generator Tube Surveillance.''

Date of issuance: November 4, 1997

Effective date: November 4, 1997, with full implementation within

30 days

Amendment Nos.: 132 and 124

Facility Operating License Nos. DPR-42 and DPR-60: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: August 13, 1997 (62 FR

43370) The August 15, September 2, and October 3, 1997, letters

provided clarifying information and updated TS pages. This information

was within the scope of the original application and did not change the

staff's initial no significant hazards considerations determination.

The Commission's related evaluation of the amendments is contained in a

Safety Evaluation dated November 4, 1997.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

PECO Energy Company, Public Service Electric and Gas Company

Delmarva Power and Light Company, and Atlantic City Electric

Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power

Station, Unit Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: June 4, 1997

Brief description of amendments: The proposed change revises the

Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, Technical

Specifications to eliminate an inconsistency between emergency core

cooling system (ECCS) operability requirements and the auto-start and

protective trip bypass of the emergency diesel generators on an ECCS

initiation signal during certain plant configurations.

Date of issuance: October 24, 1997

Effective date: Both units, as of date of issuance, to be

implemented within 30.

Amendments Nos.: 221 and 226

Facility Operating License Nos. DPR-44 and DPR-56: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: August 13, 1997 (62 FR

43373) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 24, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

PA 17105

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: May 19, 1997, as supplemented by

letter dated August 25, 1997

Brief description of amendment: This amendment changes the Hope

Creek Technical Specification (TS) 3.7.1.3, ``Ultimate Heat Sink,'' to

raise the minimum allowable ulimate heat sink (UHS) water level from 76

feet to 80 feet, lower the maximum allowable UHS temperature from

88.6 deg.F to 85 deg.F, and reflect that continued plant operation to a

UHS temperature of 87 deg.F depends upon the association of UHS

temperature and safety system availability. The associated Surveillance

Requirement, TS 4.7.1.3, is changed to decrease the river water

temperature, at which increasing temperature surveillance is required,

from 85 deg.F to 82 deg.F. The requirements of TS 3.7.1.1, ``Safety

Auxiliaries Cooling System (SACS),'' TS 3.7.1.2, ``Station Service

Water System (SSWS),'' and TS 3.8.1.1, ``Electrical Power Systems,''

are revised to reflect the revised TS 3.7.1.3. In addition, the Bases

for 3/4.7.1, ``Service Water Systems,'' are appropriately revised.

Date of issuance: October 28, 1997

Effective date: As of the date of issuance to be implemented within

60 days from the date of issuance.

Amendment No.: 106

Facility Operating License No. NPF-57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 18, 1997 (62 FR

33132) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated October 28, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: March 31, 1997, as supplemented

by letters dated July 16, August 26, and October 3, 1997

Brief description of amendment: This amendment changes Technical

Specification (TS) 2.1.2, ``THERMAL POWER, High Pressure and High

Flow,''

[[Page 61852]]

ACTION a.1.c for TS 3.4.1.1, ``Recirculation Loops,'' and the Bases for

TS 2.1, ``Safety Limits.'' These changes are being made to implement an

appropriately conservative Safety Limit Minimum Critical Power Ratio to

include Cycle 8 specific analyses for all Hope Creek core and fuel

designs.

Date of issuance: November 4, 1997

Effective date: The license amendment is effective as of its date

of issuance and shall be implemented within 60 days.

Amendment No.: 107

Facility Operating License No. NPF-57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 13, 1997 (62 FR

43374) The August 26 and October 3, 1997, letters provided clarifying

information that did not change the initial proposed no significant

hazards consideration determination. The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

November 4, 1997. No significant hazards consideration comments

received: No.

Local Public Document Room location:Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: September 29, 1997

Brief description of amendment: This amendment changes Technical

Specification (TS) 3/4.11.1, ``Liquid Effluent - Concentration.'' The

change adds a requirement to perform weekly sampling and monthly and

quarterly composite analyses of the Station Service Water System when

the Reactor Auxiliaries Cooling System is contaminated. The licensee

has also proposed an editorial change to TS Table 4.11.1.1.1-1. In

Liquid Release Type B, the licensee is proposing that the acronym for

Station Service Water System be changed from GSW to SSWS. This proposed

change will be addressed in a future license amendment.

Date of issuance: November 6, 1997

Effective date: As of date of issuance, to be implemented within 60

days.

Amendment No.: 108

Facility Operating License No. NPF-57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 6, 1997 (62 FR

52161) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 6, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: September 24, 1997

Brief description of amendment: This amendment adds a Surveillance

Requirement to Technical Specification 3/4.5.1, ``Emergency Core

Cooling Systems'', to perform a monthly valve position verification for

the four residual heat removal cross-tie valves.

Date of issuance: November 6, 1997

Effective date: As of the date of issuance, to be implemented

within 60 days.

Amendment No.: 109

Facility Operating License No. NPF-57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 6, 1997 (62 FR

52162) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 6, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: August 20, 1997

Brief description of amendment: This amendment changes the

Technical Specifications (TSs) to provide for: 1) the relocation of

suppression chamber volume references in Limiting Condition for

Operation (LCO) 3.5.3 to the Hope Creek (HC) Updated Final Safety

Analysis Report (UFSAR) and TS Bases as appropriate; 2) the revision of

the suppression chamber volume currently listed in LCO 3.5.3.b; 3) the

relocation of the suppression chamber volume references in LCO

3.6.2.1.a.1 to the UFSAR and TS Bases; and 4) the revision to the

suppression chamber volume reference in TS 5.2.1 to reference the TS

Bases section where this information will reside.

Date of issuance: November 6, 1997

Effective date: As of date of issuance, to be implemented within 60

days.

Amendment No.: 110

Facility Operating License No. NPF-57: This amendment revised the

Technical Specifications and License.

Date of initial notice in Federal Register: September 24, 1997 (62

FR 50010) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 6, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.

Ginna Nuclear Power Plant, Wayne County, New York

Date of application for amendment: August 19, 1997, as supplemented

September 29, 1997.

Brief description of amendment: The proposed amendment revises the

Ginna Station Improved Technical Specifications by adding a note to the

Containment Spray (CS) Limiting Condition for Operation 3.6.6 which

allows the CS pumps in MODE 4 to be placed in pull-stop, and motor-

operated valves 896A and 896B to have their DC control power restored

with the valves placed in the closed position in order to perform

interlock and valve testing of MOVs 857A, 857B, and 857C. A time limit

of 2 hours is placed on this configuration for each test.

Date of issuance: October 29, 1997

Effective date: October 29, 1997

Amendment No.: 68

Facility Operating License No. DPR-18: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 24, 1997 (62

FR 50011) The September 29, 1997, letter provided clarifying

information that did not change the initial proposed no significant

hazards consideration determination. The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

October 29, 1997. No significant hazards consideration comments

received: No.

Local Public Document Room location: Rochester Public Library, 115

South Avenue, Rochester, New York 14610

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and

50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston

County, Alabama

Date of amendments request: September 17, 1997

Brief description of amendments: The amendments change Technical

Specification 3/4.4.9, ``Specific Activity,'' and the associated Bases

to reduce the limit associated with dose equivalent iodine-131. The

steady-state dose equivalent iodine-131 limit would

[[Page 61853]]

be reduced by 50 percent from 0.3 mu Curie/gram to 0.15 mu Curie/gram

and the maximum instantaneous value would be reduced by 50 percent from

18 mu Curie/gram to 9 mu Curie/gram.

Date of issuance: October 29, 1997

Effective date: As of the date of issuance to be implemented within

30 days

Amendment Nos.: Unit 1 - 132; Unit 2 - 124

Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise

the Technical Specifications.

Date of initial notice in Federal Register: September 24, 1997 (62

FR 49998) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 29, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Houston-Love Memorial Library,

212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, OES Nuclear,

Inc., Pennsylvania Power Company, Toledo Edison Company, Docket No.

50-440 Perry Nuclear Power Plant, Unit 1, Lake County, Ohio

Date of application for amendment: August 14, 1997, as supplemented

September 26 and October 1, 1997.

Brief description of amendment: This amendment changes the design

basis as described in the Updated Safety Analysis Report by adding a

description of the methodology utilized for determining the systems and

components that are considered to require protection from tornado

missiles.

Date of issuance: November 4, 1997

Effective date: November 4, 1997

Amendment No.: 90

Facility Operating License No. NPF-58: This amendment revised the

license.

Date of initial notice in Federal Register: September 16, 1997 (62

FR 48674). The September 26 and October 1, 1997, submittals provided

supplemental information that did not change the initial proposed no

significant hazards consideration determination. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated November 4, 1997. No significant hazards consideration comments

received: No.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak

Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: May 16, 1997 (TXX-97119)

Brief description of amendments: The amendments revised core safety

limit curves and Overtemperature N-16 reactor trip setpoints based on

analyses of the core configuration for CPSES Unit 2, Cycle 4. These

changes apply equally to CPSES Units 1 and 2 licenses since the

Technical Specifications are combined.

Date of issuance: October 30, 1997

Effective date: October 30, 1997

Amendment Nos.: Unit 1 - Amendment No. 55; Unit 2 - Amendment No.

41

Facility Operating License Nos. NPF-87 and NPF-89: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: July 16, 1997 (62 FR

38140) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated October 30, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, TX 76019

Notice Of Issuance Of Amendments To Facility Operating Licenses And

Final Determination Of No Significant Hazards Consideration And

Opportunity For A Hearing (Exigent Public Announcement Or Emergency

Circumstances)

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application for the

amendment complies with the standards and requirements of the Atomic

Energy Act of 1954, as amended (the Act), and the Commission's rules

and regulations. The Commission has made appropriate findings as

required by the Act and the Commission's rules and regulations in 10

CFR Chapter I, which are set forth in the license amendment.

Because of exigent or emergency circumstances associated with the

date the amendment was needed, there was not time for the Commission to

publish, for public comment before issuance, its usual 30-day Notice of

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

used local media to provide notice to the public in the area

surrounding a licensee's facility of the licensee's application and of

the Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

the public to comment, using its best efforts to make available to the

public means of communication for the public to respond quickly, and in

the case of telephone comments, the comments have been recorded or

transcribed as appropriate and the licensee has been informed of the

public comments.

In circumstances where failure to act in a timely way would have

resulted, for example, in derating or shutdown of a nuclear power plant

or in prevention of either resumption of operation or of increase in

power output up to the plant's licensed power level, the Commission may

not have had an opportunity to provide for public comment on its no

significant hazards consideration determination. In such case, the

license amendment has been issued without opportunity for comment. If

there has been some time for public comment but less than 30 days, the

Commission may provide an opportunity for public comment. If comments

have been requested, it is so stated. In either event, the State has

been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an

amendment immediately effective, notwithstanding the pendency before it

of a request for a hearing from any person, in advance of the holding

and completion of any required hearing, where it has determined that no

significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has

made a final determination that the amendment involves no significant

hazards consideration. The basis for this determination is contained in

the documents related to this action. Accordingly, the amendments have

been issued and made effective as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment

[[Page 61854]]

under the special circumstances provision in 10 CFR 51.12(b) and has

made a determination based on that assessment, it is so indicated.

For further details with respect to the action see (1) the

application for amendment, (2) the amendment to Facility Operating

License, and (3) the Commission's related letter, Safety Evaluation

and/or Environmental Assessment, as indicated. All of these items are

available for public inspection at the Commission's Public Document

Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at

the local public document room for the particular facility involved.

The Commission is also offering an opportunity for a hearing with

respect to the issuance of the amendment. By December 19, 1997, the

licensee may file a request for a hearing with respect to issuance of

the amendment to the subject facility operating license and any person

whose interest may be affected by this proceeding and who wishes to

participate as a party in the proceeding must file a written request

for a hearing and a petition for leave to intervene. Requests for a

hearing and a petition for leave to intervene shall be filed in

accordance with the Commission's ``Rules of Practice for Domestic

Licensing Proceedings'' in 10 CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714 which is available at the

Commission's Public Document Room, the Gelman Building, 2120 L Street,

NW., Washington, DC and at the local public document room for the

particular facility involved. If a request for a hearing or petition

for leave to intervene is filed by the above date, the Commission or an

Atomic Safety and Licensing Board, designated by the Commission or by

the Chairman of the Atomic Safety and Licensing Board Panel, will rule

on the request and/or petition; and the Secretary or the designated

Atomic Safety and Licensing Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene

shall set forth with particularity the interest of the petitioner in

the proceeding, and how that interest may be affected by the results of

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

following factors: (1) the nature of the petitioner's right under the

Act to be made a party to the proceeding; (2) the nature and extent of

the petitioner's property, financial, or other interest in the

proceeding; and (3) the possible effect of any order which may be

entered in the proceeding on the petitioner's interest. The petition

should also identify the specific aspect(s) of the subject matter of

the proceeding as to which petitioner wishes to intervene. Any person

who has filed a petition for leave to intervene or who has been

admitted as a party may amend the petition without requesting leave of

the Board up to 15 days prior to the first prehearing conference

scheduled in the proceeding, but such an amended petition must satisfy

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

scheduled in the proceeding, a petitioner shall file a supplement to

the petition to intervene which must include a list of the contentions

which are sought to be litigated in the matter. Each contention must

consist of a specific statement of the issue of law or fact to be

raised or controverted. In addition, the petitioner shall provide a

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

contention and on which the petitioner intends to rely in proving the

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

aware and on which the petitioner intends to rely to establish those

facts or expert opinion. Petitioner must provide sufficient information

to show that a genuine dispute exists with the applicant on a material

issue of law or fact. Contentions shall be limited to matters within

the scope of the amendment under consideration. The contention must be

one which, if proven, would entitle the petitioner to relief. A

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses. Since the Commission has made a final determination

that the amendment involves no significant hazards consideration, if a

hearing is requested, it will not stay the effectiveness of the

amendment. Any hearing held would take place while the amendment is in

effect.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555-0001, Attention: Rulemakings and

Adjudications Staff or may be delivered to the Commission's Public

Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,

by the above date. A copy of the petition should also be sent to the

Office of the General Counsel, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for a hearing will

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

petition and/or request should be granted based upon a balancing of the

factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of application for amendment: October 24, 1997

Brief description of amendment: The amendment adds a footnote to

Technical Specification 3.7.A.5, ``Primary Containment.'' The footnote

provides a one time exception to the reverse flow testing requirement

for containment isolation check valve 30-CK-432.

Date of issuance: October 30, 1997

Effective date: As of date of issuance and shall be implemented by

November 2, 1997.

Amendment No.: 174

Facility Operating License No. DPR-35: This amendment revised the

Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration: No.

The Commission's related evaluation of the amendment, finding of

emergency circumstances, consultation with the State of Massachusetts,

and final no significant hazards consideration determination are

contained in a Safety Evaluation dated October 30, 1997.

Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,

800 Boylston Street, 36th Floor, Boston, Massachusetts 02199

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360

NRC Project Director: Ronald B. Eaton, Acting Director

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey County, Kansas

Date of amendment request: October 17, 1997

[[Page 61855]]

Brief description of amendment: The amendment revised Technical

Specification 4.5.2b and associated Bases to eliminate the requirement

to vent the centrifugal charging pump casings.

Date of issuance: November 3, 1997

Effective date: November 3, 1997

Amendment No.: 114

Facility Operating License No. NPF-42: The amendment revised the

Technical Specifications. Press release issued requesting comments as

to proposed no significant hazards consideration: Yes. October 24,

1997. Coffey County Today Newspaper (Kansas). Comments received: Yes.

Comments were submitted by Mr. Dave Lochbaum of the Union of Concerned

Scientists by letter dated October 29, 1997. Verbal comments were

received from Larry Myers on October 28, 1997. The staff responded to

these comments in the safety evaluation attached to the November 3,

1997, amendment. The Commission's related evaluation of the amendment,

finding of exigent circumstances, consultation with the State of Kansas

and final determination of no significant hazards consideration are

contained in a Safety Evaluation dated November 3, 1997.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037

NRC Project Director: William H. Bateman

Dated at Rockville, Maryland, this 12th day of November 1997.

For the Nuclear Regulatory Commission

Elinor G. Adensam,

Acting Director, Division of Reactor Projects - III/IV Office of

Nuclear Reactor Regulation.

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

BILLING CODE 7590-01-F