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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

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

[[Page 6969]]

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 January 16, 1998, through January 30, 1998.

The last biweekly notice was published on January 28, 1998 (63 FR

4308).

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed no Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Administration Services, Office of

Administration, U.S. Nuclear Regulatory Commission, Washington, DC

20555-0001, and should cite the publication date and page number of

this Federal Register notice. Written comments may also be delivered to

Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,

Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of

written comments received may be examined at the NRC Public 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 March 13, 1998, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

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significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of amendment request: November 1, 1996.

Description of amendment request: The change increases the

surveillance interval to allow verification that a reactivity anomaly

does not exist to every 1100 MWD/T (megawatt-days per metric ton)

average core exposure (approximately 41 days) instead of once every one

effective full power month (approximately 30 days).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

This change increases the surveillance interval to allow

verification that a reactivity anomaly does not exist every 1100 MWD/T

average core exposure (approximately 41 days) instead of once every one

effective full power month (approximately 30 days). Reactivity

anomalies are not considered to be initiators of any analyzed event.

Operating history has shown that the difference between predicted and

monitored core reactivity is continually acceptable during the extended

Surveillance interval. The consequences of an accident are not affected

by relaxing the Frequency of the Surveillance since the consequences of

an event with a reactivity anomaly during the current interval (due to

not detecting the existence of a reactivity anomaly between

Surveillances) are the same as the consequences of an event with a

reactivity anomaly during the additional period. Additionally, the most

common outcome of the performance of a Surveillance is the successful

demonstration that the acceptance criteria are satisfied. This change

does not alter assumptions relative to the mitigation of an accident or

transient event. Therefore, this change does not involve a significant

increase in the probability or consequences of a previously analyzed

accident.

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

kind of accident from any accident previously evaluated?

The change introduces no new mode of plant operation and it does

not involve physical modification to the plant. Therefore, it does not

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

accident previously evaluated.

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

safety?

The proposed change is acceptable since the proposed Frequency is

adequate for ensuring a reactivity anomaly does not exist. Operating

history has shown that the difference between predicted and monitored

core reactivity is continually acceptable during the extended

Surveillance interval. Also, this change is considered acceptable since

the most common outcome of the performance of a Surveillance is the

successful demonstration that the acceptance criteria are satisfied.

The safety analysis assumptions will still be maintained, thus, no

question of safety exists. Therefore, this change does not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of amendment request: November 1, 1996.

Description of amendment request: The current Technical

Specifications (TS) for the Brunswick Steam Electric Plant (BSEP) only

address a single inoperable scram accumulator, requiring entry into TS

3.0.3 for direction to shut down a unit if additional scram

accumulators become inoperable. The proposed change corrects this

situation by revising the declared status of control rods with

inoperable scram accumulators and allowing a short out-of-service time

for the control rod scram accumulators before requiring a unit

shutdown, consistent with the Improved Technical Specifications (ITS)

(NUREG-1433, ``Standard Technical Specifications General Electric

Plants, BWR/4,'' Revision 1, April 1995). In the event scram

accumulators are inoperable concurrent with low charging water header

pressure, the ITS require that the reactor mode switch be placed in the

``shutdown'' position, which ensures that all control rods are inserted

and the unit is shutdown. The proposed change deviates from the ITS in

that it requires a manual scram under these conditions which also

ensures that all control rods are inserted and the unit is shutdown.

Details associated with this deviation are included in a Carolina Power

& Light Company letter dated September 11, 1997 (see response to NRC

comment 3.1.5-2), which is available to the public.

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:

[[Page 6971]]

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change revises the declared status of control rods

with inoperable scram accumulators and allows a short out-of-service

time for the control rod scram accumulators before requiring a plant

shutdown. Inoperable scram accumulators are not considered initiators

for any accidents previously evaluated, and therefore, cannot increase

the probability of such accidents. The extended time period to declare

a control rod inoperable provides a reasonable time to attempt

investigation and restoration of the inoperable control rod scram

accumulator. This time period is acceptable since the time period is

sufficiently short such that it does not increase the risk significance

of an ATWS [anticipated transient without scram] event. Furthermore,

this change will add actions which will address the situation where

multiple control rod scram accumulators may rapidly become inoperable.

In addition, the change that allows modifying the status of a control

rod with an inoperable scram accumulator is acceptable since the

numbers and distribution of control rods are restricted and Technical

Specification actions continue to ensure that the control rods can

still perform their safety function when required. As a result, this

change will not involve a significant increase in the consequences of

an accident previously evaluated.

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 physical modification to the

plant. The change in the operation is consistent with current safety

analysis assumptions. Therefore, the change does not create the

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

previously evaluated.

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

safety?

The proposed change is consistent with the assumptions of the

current safety analysis. The extended time to evaluate and access two

or more inoperable control rod scram accumulators and the allowance to

declare any control rod with an inoperable scram accumulator ``slow''

when operating at a reactor pressure [greater than or equal to] 950

psig proposed by this change is acceptable since adequate controls are

added to the Technical Specifications which ensure charging water

header pressure to the control rod scram accumulators is maintained and

action is provided to immediately shutdown the reactor before the scram

safety function is significantly impacted in the event cha[r]ging water

header pressure cannot be maintained. Therefore, the proposed change

does not involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina.

Date of amendment request: November 1, 1996.

Description of amendment request: The proposed changes extend the

refueling interval surveillance Frequencies that are currently

specified as 18 months for surveillances other than those associated

with instrumentation channel calibration to 24 months.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes involve a change in the surveillance Frequency

from 18 months to 24 months. The change in surveillance Frequency is

not assumed to be an accident initiator for any accidents previously

evaluated in the SAR [Updated Final Safety Analysis Report]. Therefore,

this change will have no impact on the probability of an accident

previously evaluated. By changing the Surveillance Frequency from 18

months plus grace to a maximum of 30 months, the consequences of an

accident previously evaluated in the SAR are not significantly

increased. This is based on the fact that the evaluation of the subject

changes demonstrated that the overall impact, if any, on the systems[']

availability is minimal. Since the impact on the systems is minimal, it

can be concluded that the overall impact on the plant accident analysis

is negligible. Furthermore, it is shown that the performance history

for the subject systems does not indicate any failures which would

invalidate the conclusions reached in this evaluation.

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

kind of accident from any accident previously evaluated?

This proposed change will not involve any physical changes to plant

systems, structures, or components. The changes in normal plant

operation are consistent with the current safety analysis assumptions.

Therefore, this change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

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

safety?

The margin of safety has not been significantly reduced. Although,

there will be an increase in the interval between the subject

surveillance tests, the evaluation of the changes demonstrates that

there is no evidence of any failures which would impact the subject

systems['] availability. Based on the fact that the increased testing

interval has a minimal impact on the subject systems, it can be

concluded that the assumptions in the licensing basis are not impacted

by the changes in the subject requirements and commitments.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

[[Page 6972]]

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of amendment request: November 1, 1996.

Description of amendment request: The proposed change involves a

change in the instrumentation channel calibration surveillance testing

intervals from 18 months to 24 months.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change involves a change in the instrumentation

channel calibration surveillance testing intervals from 18 months to 24

months. The proposed change does not physically impact the plant nor

does it impact any design or functional requirements of the associated

systems. That is, the proposed change does not degrade the performance

or increase the challenges of any safety systems assumed to function in

the accident analysis. The proposed change does not impact the

Surveillance Requirements themselves nor the way in which the

Surveillances are performed. Additionally, the proposed change does not

introduce any new accident initiators since no accidents previously

evaluated have as their initiators anything related to the frequency of

surveillance testing. The proposed change does not affect the

availability of equipment or systems required to mitigate the

consequences of an accident because of the availability of redundant

systems or equipment and because other test[s] performed more

frequently will identify potential equipment problems. Furthermore, a

historical review of surveillance test results indicated that all

failures identified were unique, non-repetitive, and not related to any

time-based failure modes, and indicated no evidence of any failures

that would invalidate the above conclusions. Therefore, the proposed

change does not increase the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change involves a change in the instrumentation

channel calibration surveillance testing intervals from 18 months to 24

months. The proposed change does not introduce any failure mechanisms

of a different type than those previously evaluated since there are no

physical changes being made to the facility. In addition, the

Surveillance Requirements themselves and the way Surveillances are

performed will remain unchanged. Furthermore, a historical review of

surveillance test results indicated no evidence of any failures that

would invalidate the above conclusions. Therefore, the proposed change

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

from any previously evaluated.

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

safety?

Although the proposed change will result in an increase in the

interval between surveillance tests, the impact on system availability

is small based on other, more frequent testing or redundant systems or

equipment, and there is no evidence of any failures that would impact

the availability of the systems. Therefore, the assumptions in the

licensing basis are not impacted, and the proposed change does not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of amendment request: November 1, 1996.

Description of amendment request: The proposed change allows a

short out-of-service time for various combinations of inoperable

emergency core cooling system (ECCS) subsystems instead of an immediate

plant shutdown.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change allows a short out-of-service time for various

combinations of inoperable ECCS subsystems instead of an immediate

plant shutdown. ECCS equipment is used to mitigate the consequences of

an accident, but the inoperability of ECCS equipment is not considered

as the initiator of any previously analyzed accident. As such, the

inoperability of ECCS subsystems will not increase the probability of

any accident previously evaluated. The proposed combinations of

inoperable ECCS subsystems are bounded by the analysis summarized in

NEDC-31624P which utilizes an NRC [Nuclear Regulatory Commission]

approved methodology for determining consequences. This analysis

demonstrated that adequate core cooling would still be provided with

the proposed change. Therefore, the consequences of an event occurring

during the proposed allowed outage time are the same as the

consequences of an event occurring during the current period allowed to

place the plant in a shutdown condition. As a result, the change does

not involve a significant increase in the consequences of any accident

previously evaluated.

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 introduce a new mode of plant

operation and does not involve physical modification to the plant.

Therefore, it does not create the possibility of a new or different

kind of accident from any accident previously evaluated.

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

safety?

The proposed combinations of inoperable ECCS subsystems are bounded

by the analysis summarized in NEDC-31624P which utilizes an NRC

approved methodology. This analysis demonstrated that adequate core

cooling would still be provided with the proposed change. In addition,

the allowable outage time specified is based on a reliability study

(Memorandum from R.L. Baer (NRC) to V. Stello, Jr. (NRC), ``Recommended

Interim Revisions to LCOs [limiting conditions

[[Page 6973]]

for operation] for ECCS Components,'' December 1, 1975) and has been

found to be acceptable through operating experience. Any reduction in

the margin of safety is offset by the benefit of reducing the transient

risk associated with an immediate plant shutdown. 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: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of amendment request: November 1, 1996.

Description of amendment request: The proposed change reduces the

number of automatic depressurization system (ADS) valves required to be

OPERABLE from seven to six.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change reduces the number of ADS valves required to be

OPERABLE from seven to six. The number of ADS valves required to be

OPERABLE is not assumed in the initiation of any analyzed event.

Therefore, the change does not increase the probability of an accident

previously evaluated.

The ADS valves function to mitigate the consequences of analyzed

events by reducing the reactor vessel pressure to allow low pressure

ECCS [emergency core cooling system] components to function as needed

in the event of a HPCI [high-pressure coolant injection] System

failure. The change is based on the analysis summarized in NEDC-31624P,

``Brunswick Steam Electric Plant Units 1 and 2 SAFER/GESTR-LOCA Loss-

of-Coolant Accident Analysis,'' Revision 2, July 1990. This analysis

shows that adequate core cooling is provided during a small break LOCA

and a simultaneous HPCI System failure (limiting LOCA) with two of the

seven ADS valves out-of-service. NEDC-31624P was previously reviewed

and accepted by the NRC [Nuclear Regulatory Commission] as documented

in a letter from E.G. Tourigny (NRC) to L.W. Eury (CP&L), ``SAFER/

GESTR-LOCA Analysis, Brunswick Steam Electric Plant, Units 1 and 2 (TAC

Nos. 72854/72855),'' dated 06/01/89 and a letter from E.G. Tourigny

(NRC) to L.W. Eury (CP&L), ``Revision of SAFER/GESTR-LOCA Analysis--

Brunswick Steam Electric Plant, Units 1 and 2 (TAC Nos. 77585 and

77586),'' dated 01/10/91. The change is considered acceptable since the

analyses show that only five ADS valves are required to perform the

intended safety function of lowering reactor pressure. As a result, the

change does not involve a significant increase in the consequences of

an accident previously evaluated.

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 physical modification to the

plant and the proposed change continues to provide assurance that the

ADS can perform its intended safety function when required. Therefore,

it does not create the possibility of a new or different kind of

accident from any accident previously evaluated.

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

safety?

This proposed change does not involve a significant reduction in a

margin of safety since sufficient ADS valves are maintained to ensure

the safety analysis assumptions are met. The safety analysis shows

that, with a HPCI failure, five ADS valves are sufficient to lower

reactor pressure to allow low pressure ECCS injection and cooling.

Thus, the proposed change does not impact the 10 CFR 50.46 limits.

NEDC-31624P was previously reviewed and accepted by the NRC as

documented in a letter from E.G. Tourigny (NRC) to L.W. Eury (CP&L),

``SAFER/GESTR-LOCA Analysis, Brunswick Steam Electric Plant, Units 1

and 2 (TAC Nos. 72854/72855),'' dated 06/01/89 and a letter from E.G.

Tourigny (NRC) to L.W. Eury (CP&L), ``Revision of SAFER/GESTR-LOCA

Analysis--Brunswick Steam Electric Plant, Units 1 and 2 (TAC Nos. 77585

and 77586),'' dated 01/10/91. As a result, this change does not involve

a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of amendment request: November 1, 1996.

Description of amendment request: This change will raise the

minimum pressure at which the automatic depressurization system (ADS)

is required to be OPERABLE to 150 psig.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

This change will raise the minimum pressure at which ADS is

required to be OPERABLE to 150 psig. The OPERABILITY of the ADS valves

below 150 psig is not assumed in the initiation of any analyzed event.

The ADS is assumed in the mitigation of consequences of a LOCA [loss-

of-coolant accident] which occurs at high reactor pressure. The ADS is

not assumed in the mitigation of low reactor pressure events since its

function is to lower the pressure to within the capabilities of the low

pressure makeup systems. Low pressure injection systems are analyzed

(per NEDC-31624P, ``Brunswick Steam Electric Plant Units

[[Page 6974]]

1 and 2 SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis,'' Revision

2, July 1990) to begin injection into the RPV [reactor pressure vessel]

at pressures well above 150 psig. As a result, the proposed change does

not impact the ability of the ECCS [emergency core cooling system] to

perform [its] intended safety function and the change does not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

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 physical modification to the

plant and the proposed change continues to provide assurance that the

ADS can perform its safety function when required. Therefore, the

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

kind of accident from any accident previously evaluated.

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

safety?

The purpose of the ADS is to lower reactor pressure sufficiently to

allow low pressure ECCS to inject and cool the core in the event of a

HPCI [high-pressure coolant injection] System failure. Revising the

minimum pressure for required ADS valve OPERABILITY is acceptable since

the low pressure ECCS can provide core cooling at reactor pressures

well above 150 psig and since the HPCI System is not required to be

OPERABLE below 150 psig. As a result, 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: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of amendment request: November 1, 1996.

Description of amendment request: The proposed change relaxes the

low pressure emergency core cooling system (ECCS) pump flow acceptance

criteria under operational conditions 1 (power operation), 2 (startup),

and 3 (hot shutdown).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change relaxes the low pressure ECCS pump flow

acceptance criteria. Low pressure ECCS equipment is used to mitigate

the consequences of an accident, but is not considered as the initiator

of any previously analyzed accident. As such, the change does not

increase the probability of any accident previously evaluated. The

proposed low pressure ECCS pump flow acceptance criteria are assumed in

the analysis summarized in NEDC-31624P [``Brunswick Steam Electric

Plant Units 1 and 2 SAFR/GESTR-LOCA Loss-of-Coolant Accident

Analysis,'' Revision 2, July 1990] which utilizes an NRC approved

methodology for determining consequences. The resulting peak cladding

temperature for all the cases analyzed in NEDC-31624P is below 1600

deg.F (a significant margin to the 10 CFR 50.46 limit). As a result,

the ECCS subsystems assumed to be available during events analyzed will

continue to provide adequate core cooling. Therefore, the change does

not involve a significant increase in the consequences of any accident

previously evaluated.

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 introduce a new mode of plant

operation and does not involve physical modification to the plant. In

addition, the low pressure ECCS flow rates will not be determined in a

new or different way. Therefore, it does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

safety?

The proposed low pressure ECCS pump flow acceptance criteria are

assumed in the analysis summarized in NEDC-31624P which utilizes an NRC

approved methodology. NEDC-31624P concludes that the ECCS subsystems

can still provide adequate core cooling with the proposed pump flow

acceptance criteria and in all cases analyzed peak cladding temperature

is maintained below 1600 deg.F. In addition, plant procedures will

continue to trend the performance of the low pressure ECCS pumps and

ensure that any adverse trends in equipment performance are identified

and appropriate actions taken. 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: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of amendment request: November 1, 1996.

Description of amendment request: The proposed change relaxes the

core spray (CS) pump flow acceptance criterion during shutdown

conditions.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change relaxes the CS pump flow acceptance criterion.

Low pressure ECCS [emergency core cooling

[[Page 6975]]

system] equipment is used to mitigate the consequences of a reactor

vessel draindown event during shutdown conditions, but is not

considered as the initiator of any previously analyzed accident. As

such, the change does not increase the probability of any accident

previously evaluated. The proposed low pressure ECCS pump flow

acceptance criteria are assumed in the analysis summarized in NEDC-

31624P [``Brunswick Steam Electric Plant Units 1 and 2 SAFR/GESTR-LOCA

Loss-of-Coolant Accident Analysis,'' Revision 2, July 1990] which

utilizes an NRC approved methodology for determining consequences. The

resulting peak cladding temperature for all the cases analyzed in NEDC-

31624P is below 1600 deg.F (a significant margin to the 10 CFR 50.46

limit). This analysis assumes the reactor was operating at high power.

This analysis did not invalidate the long term cooling analysis

described in NEDO-20566A [``General Electric Company Analytical Model

for Loss of Coolant Analysis in accordance with 10 CFR 50 Appendix

K'']. Therefore, since the CS pump flow proposed by this change is

adequate for high power conditions, it is reasonable to assume the CS

pump flow is adequate to restore and maintain adequate vessel level

during an inadvertent vessel draindown event while shutdown. The

required low pressure ECCS subsystems during events analyzed in

shutdown conditions will continue to provide adequate redundancy and

coolant makeup capability. Therefore, the change does not involve a

significant increase in the consequences of any accident previously

evaluated.

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 introduce a new mode of plant

operation and does not involve physical modification to the plant. In

addition, the CS pump flow rate will not be determined in a new or

different way. Therefore, it does not create the possibility of a new

or different kind of accident from any accident previously evaluated.

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

safety?

The proposed CS pump flow acceptance criterion is assumed in the

analysis summarized in NEDC-31624P which utilizes an NRC approved

methodology. NEDC-31624P concludes that the ECCS subsystems can still

provide adequate core cooling with the proposed CS pump flow acceptance

criterion and in all cases analyzed peak cladding temperature is

maintained below 1600 deg.F. Since the analysis assumed high power

conditions, it is reasonable to assume that, with the proposed change,

adequate coolant makeup capability is maintained during shutdown

conditions. In addition, plant procedures will continue to trend the

performance of the low pressure ECCS pumps and ensure that any adverse

trends in equipment performance are identified and appropriate actions

taken. 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: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of amendment request: November 1, 1996.

Description of amendment request: This proposed change eliminates

current Technical Specification (CTS) 3/4.6.1.5, Primary Containment

Internal Pressure.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

This proposed change eliminates CTS 3/4.6.1.5, Primary Containment

Internal Pressure. This change does not result in any hardware or

operating procedure changes. The primary containment pressure is not

assumed to be an initiator of any analyzed event. It is an initial

condition in the containment analysis (e.g., following a DBA LOCA

[design-basis accident loss-of-coolant accident]). CTS 3/4.6.1.5 was

necessary to maintain this assumption which helps ensure that the

primary containment design pressure is not exceeded following an

accident. However, the power uprate analysis modified this initial

drywell pressure value such that the assumed value is greater than the

RPS [reactor protection system] high drywell trip. The results of the

power uprate analysis show that this modified initial drywell pressure

is acceptable for ensuring primary containment pressure design limits

are not exceeded. This modified initial pressure was utilized in

determining a new Pa [calculated peak containment internal

pressure related to the design basis accident], and has been submitted

to the NRC to support the BNP [Brunswick Nuclear Plant] power uprate

amendment.

The initial drywell pressure assumption is being ensured by the RPS

high drywell pressure scram, which will trip the unit prior to

exceeding the assumed drywell pressure value, effectively placing the

unit in MODE 3. While the RPS trip is not required in MODE 3, the

Emergency Operating Procedures (EOPs) will govern actions if the

drywell pressure exceeds the assumed drywell pressure value. The EOPs

will require entry into the Reactor Vessel Control and Primary

Containment Control actions. These actions require steps to reduce

primary containment pressure to below the value assumed in the accident

analyses and to cool down the reactor at normal cooldown rates to MODE

4 if pressure cannot be reduced below the reactor trip setpoint. The

negative pressure limit is controlled and met by the design and proper

operation of the reactor building-to-suppression chamber and the

suppression chamber-to-drywell vacuum breakers. These vacuum breakers,

which are required to be OPERABLE in MODES 1, 2, and 3, are designed to

ensure the negative pressure design limit of the primary containment is

not exceeded. Therefore, this change will not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

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 introduce a new mode of plant

operation and does not require physical modification to the plant.

Therefore, the change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

[[Page 6976]]

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

safety?

No significant reduction in a margin of safety is involved. The

upper pressure limit is maintained by the design and proper operation

of the RPS high drywell pressure trip, a Technical Specification

required instrumentation function, and the EOPs. The negative pressure

limit is being maintained by the design and proper operation of the

reactor building-to-suppression chamber and suppression chamber-to-

drywell vacuum breakers, also Technical Specification required

components. Therefore, adequate controls exist with respect to the

primary containment pressure limits to ensure the primary containment

pressure will not be exceeded in the event of a design basis event.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of amendment request: November 1, 1996.

Description of amendment request: The proposed change relocates

requirements and surveillances for the Containment Air Dilution (CAD)

system from the Technical Specifications to a licensee controlled

document. Licensee analysis has demonstrated that the CAD system is not

needed to maintain the primary containment atmosphere below

flammability limits.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change relocates requirements and surveillances for

structures, systems, components or variables that do not meet the

criteria for inclusion in Technical Specifications as identified in the

Application of Selection Criteria to the BNP [Brunswick Nuclear Plant]

Technical Specifications. The affected structures, systems, components

or variables are not assumed to be initiators of analyzed events and

are not assumed to mitigate accident or transient events. The

requirements and surveillances for these affected structures, systems,

components or variables will be relocated from the Technical

Specifications to an appropriate administratively controlled document

which will be maintained pursuant to 10 CFR 50.59. In addition, the

affected structures, systems, components or variables are addressed in

existing surveillance procedures which are also controlled by 10 CFR

50.59 and subject to the change control provisions imposed by plant

administrative procedures, which endorse applicable regulations and

standards. Therefore, this change does not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

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 a physical alteration of the

plant (no new or different type of equipment will be installed) or a

change in the methods governing normal plant operation. The proposed

change will not impose or eliminate any requirements and adequate

control of existing requirements will be maintained. Thus, this change

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

from any accident previously evaluated.

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

safety?

The proposed change will not reduce a margin of safety because it

has no impact on any safety analysis assumptions. In addition, the

relocated requirements and surveillances for the affected structure,

system, component or variable remain the same as the existing Technical

Specifications. Since any future changes to these requirements or the

surveillance procedures will be evaluated per the requirements of 10

CFR 50.59, no reduction in a margin of safety will be permitted.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of amendment request: November 1, 1996.

Description of amendment request: The proposed change applies to

the Brunswick Steam Electric Plant (BSEP), Units 1 and 2, and provides

longer out-of-service times for various combinations of inoperable

service water (SW) pumps and deletes various limitations of which pumps

can be inoperable.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change provides longer out-of-service times for

various combinations of inoperable SW pumps and deletes various

limitations of which pumps can be inoperable (e.g., a remaining unit

specific NSW [nuclear service water] pump must be electrically

separated from the remaining CSW [conventional service water] pump).

The SW System supports safety related systems used to mitigate the

consequences of an accident, but the inoperability of the SW System is

not considered as the initiator of any previously analyzed accident. As

such, the inoperability of SW pumps will not increase the probability

of any accident previously evaluated. The proposed

[[Page 6977]]

combinations of inoperable SW pumps are bounded by the analyses

summarized in CP&L calculations PCN GOO50A-10 [``BSEP Unit No. 1

Service Water System Hydraulic Analysis,'' Revision 6, dated July 29,

1993] and PCN GOO50A-12 [``BSEP Unit No. 2 Service Water System

Hydraulic Analysis,'' Revision 5, dated August 11, 1992] which have

been previously evaluated by the NRC. These analyses demonstrate that

adequate SW cooling capability would still be provided with the

proposed changes. Therefore, the consequences of an event occurring

during the proposed allowed outage times are the same as the

consequences of an event occurring during the current allowed outage

time period or the current period allowed to place the plant in a

shutdown condition. As a result, the change does not involve a

significant increase in the consequences of any accident previously

evaluated.

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 physical modification to the

plant or changes in parameters governing normal plant operation. The

proposed change continues to provide assurance that the SW System is

capable of performing its required support function. Therefore, the

change does not create the possibility of a new or different kind of

accident from any accident previously evaluated.

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

safety?

The proposed combinations of inoperable SW pumps are bounded by the

analyses summarized in CP&L calculations PCN GOO50A-10 and PCN GOO50A-

12 which have been previously evaluated by the NRC. These analyses

demonstrate that adequate SW cooling capability would still be provided

with the proposed change. In addition, the proposed allowable outage

times and the capability of the SW System to support additional single

failures are consistent with the allowable outage times and capability

of other safety related systems with similar levels of degradation. Any

reduction in the margin of safety is offset by the benefit of reducing

the transient risk associated with an unnecessary plant shutdown.

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: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of amendment request: November 1, 1996.

Description of amendment request: The proposed change allows the

extension of the Allowed Outage Time (AOT) from 24 hours to 7 days of a

shutdown unit's 4.16 kilovolt (kV) balance of plant (BOP) bus which is

needed to support loads required by the operating unit.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

Extending AOT of a shutdown unit's BOP bus from 24 hours to 7 days

will not increase the probability of occurrence of an accident on the

operating unit. The probability of a previously evaluated accident

would not be increased by the longer AOT since de-energization of a

single BOP bus is not considered in the initiation of any previously

analyzed event. The BOP buses support the distribution of offsite power

to the Class 1E AC Electrical Power Distribution System, which supports

equipment necessary for the mitigation of accidents. Extending the AOT

of a shutdown unit's BOP bus will not significantly increase the

consequences of an accident on the operating unit. The consequences of

an accident occurring during the proposed 7 day AOT would be the same

as the consequences associated with the existing 24 hour AOT.

Therefore, this change will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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 introduce a new mode of plant

operation and does not involve a physical modification to the plant.

Therefore, it does not create the possibility of a new or different

kind of accident from any accident previously evaluated.

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

safety?

The margin of safety is defined by the scenario where a LOCA [loss-

of-coolant accident] occurs on the operating unit concurrent with loss

of offsite power and the worst case single failure (e.g., loss of a DG

[diesel generator] and associated supported loads). The intentional de-

energization of one of the AC Electrical Power Distribution System load

groups primarily associated with the shutdown unit, as a result of de-

energization of a BOP bus associated with the shutdown unit, will leave

three AC Electrical Power Distribution System load groups OPERABLE each

with their associated emergency diesel generator and two sources of

offsite power OPERABLE. Two of these AC Electrical Power Distribution

System load groups will be associated with the operating unit and one

with the shutdown unit. Loss of an AC Electrical Power Distribution

System load group primarily associated with the shutdown unit is not as

limiting to the operating unit as the loss of one of its emergency

power system load groups; there are fewer operating unit loads required

for mitigation of accident and transients affected by the removal of an

AC Electrical Power Distribution System load group primarily associated

with the shutdown unit. The intentional de-energization of an AC

Electrical Power Distribution System load group primarily associated

with the shutdown unit, as a result of de-energization of a BOP bus, is

enveloped by the LOCA scenario described above.

There are a number of operating unit loads required for mitigation

of accidents and transients which will become inoperable when an AC

Electrical Power Distribution System load group primarily associated

with the shutdown unit is removed from service as a result of de-

energization of the associated BOP bus. A review of the loads supported

by each of the load groups indicates that operating unit loads required

for mitigation of accidents and transients can either be

[[Page 6978]]

supplied from an alternate source or the Technical Specifications would

allow an AOT of 7 days or greater for the affected loads. Changing the

AOT from 24 hours to 7 days for an inoperable BOP bus associated with

the shutdown unit would not exceed the AOT for these individual loads.

In addition, operating unit primary containment isolation valves

supplied from the shutdown unit's out of service load group (RHR

[residual heat removal] Outboard Injection, RHR Inboard Injection, and

RHR Torus Spray) would be closed, in accordance with the Technical

Specification requirements of the operating unit, to ensure they

perform their safety function if needed. The proposed AOT for an

inoperable BOP bus associated with [the] shutdown unit provides the

benefit of improved reliability and availability of the AC Electrical

Power Distribution System and the associated offsite power circuits

(via upstream BOP buses) since the longer AOT will allow maintenance of

the buses of these load groups to be performed on a more optimum

schedule. As a result, the proposed change does not involve a

significant decrease in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of amendment request: November 1, 1996.

Description of amendment request: The proposed change allows

extension of the Allowed Outage Time (AOT) from 8 hours to 7 days of

one of the shutdown unit's emergency load groups which is needed to

support loads required by the operating unit.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

Extending the Allowed Outage Time (AOT) of an AC Electrical Power

Distribution System load group primarily associated with a shutdown

unit from 8 hours to 7 days will not increase the probability of

occurrence of an accident on the operating unit. The probability of a

previously evaluated accident would not be increased by the longer AOT

since de-energization of a single load group is not considered in the

initiation of any previously analyzed event. The Class 1E AC Electrical

Power Distribution System supports equipment necessary for the

mitigation of accidents. Extending the AOT of an AC Electrical Power

Distribution System load group primarily associated with a shutdown

unit will not significantly increase the consequences of an accident on

the operating unit. The consequences of an accident occurring during

the proposed 7 day AOT would be the same as the consequences associated

with the existing 8 hour AOT. Therefore, this change will not involve a

significant increase in the probability or consequences of an accident

previously evaluated.

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 introduce a new mode of plant

operation and does not involve a physical modification to the plant.

Therefore, it does not create the possibility of a new or different

kind of accident from any accident previously evaluated.

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

safety?

The margin of safety is defined by the scenario where a LOCA [loss-

of-coolant] occurs on the operating unit concurrent with loss of

offsite power and the worst case single failure (e.g., loss of a DG

[diesel generator] and associated supported loads). The intentional de-

energization of one of the AC Electrical Power Distribution System load

groups primarily associated with the shutdown unit will leave three AC

Electrical Power Distribution System load groups OPERABLE each with

their associated emergency diesel generator and two sources of offsite

power OPERABLE. Two of these AC Electrical Power Distribution System

load groups will be associated with the operating unit and one with the

shutdown unit. Loss of an AC Electrical Power Distribution System load

group primarily associated with the shutdown unit is not as limiting to

the operating unit as the loss of one of its emergency power system

load groups; there are fewer operating unit loads required for

mitigation of accident and transients affected by the removal of an AC

Electrical Power Distribution System load group primarily associated

with the shutdown unit. The intentional de-energization of an AC

Electrical Power Distribution System load group primarily associated

with the shutdown unit is enveloped by the LOCA scenario described

above.

There are a number of operating unit loads required for mitigation

of accidents and transients which will become inoperable when an AC

Electrical Power Distribution System load group primarily associated

with the shutdown unit is removed from service. A review of the loads

supported by each of the load groups indicates that operating unit

loads required for mitigation of accidents and transients can either be

supplied from an alternate source or the Technical Specifications would

allow an AOT of 7 days or greater for the affected loads. Changing the

AOT from 8 hours to 7 days for an inoperable AC Electrical Power

Distribution System load group primarily associated with a shutdown

unit would not exceed the AOT for these individual loads. In addition,

operating unit primary containment isolation valves supplied from the

shutdown unit's out of service load group (RHR [residual heat removal]

Outboard Injection, RHR Inboard Injection, and RHR Torus Spray) would

be closed, in accordance with the Technical Specification requirements

of the operating unit, to ensure they perform their safety function if

needed. The proposed AOT for an inoperable AC Electrical Power

Distribution System load group provides the benefit of improved

reliability and availability of the AC Electrical Power Distribution

System since the longer AOT will allow maintenance of the buses of

these load groups to be performed on a more optimum schedule. As a

result, the proposed change does not involve a significant decrease 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

[[Page 6979]]

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of amendment request: November 1, 1996.

Description of amendment request: The proposed change allows

reactor coolant system (RCS) hydrostatic pressure and leakage testing

to be performed with average reactor coolant temperature in excess of

212 deg.F and not consider the plant to be in MODE 3 (hot shutdown)

provided certain conditions are met.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated? The

proposed change allows RCS hydrostatic pressure and leakage testing to

be performed with average reactor coolant temperature in excess of

212 deg.F and not consider the plant to be in MODE 3 provided certain

conditions are met. The probability of a leak or a pipe break in the

reactor coolant pressure boundary during inservice leak and hydrostatic

testing is not increased by allowing reactor coolant temperature to

exceed 212 deg.F because the Reactor Coolant System is designed for

temperatures exceeding 500 deg.F with similar pressures. In addition,

because an inspection is being performed on the Reactor Coolant System

piping while it is being pressurized, the probability of a crack going

unnoticed and resulting in a pipe break is reduced. Reactor vessel

integrity will not be compromised by performing hydrostatic pressure

and leakage testing at temperatures in excess of 212 deg.F. Performing

hydrostatic pressure and leakage testing above 212 deg.F would allow

steam, rather than water to emit from a leak or pipe break. The

hydrostatic or inservice leak test is performed with a water solid

reactor pressure vessel. An engineering analysis was performed to

determine the reactor building pressure and temperature effects if a

pipe break occurred during the hydrostatic pressure and inservice leak

testing at a reactor coolant temperature of 275 deg.F. A recirculation

line break was used in the analysis since it was considered the most

conservative pipe break with primary containment breached during the

test. This analysis has concluded that the recirculation line break

during the performance of the test could result in a rise in reactor

building pressure sufficient to cause the opening of the reactor

building blowout panel and result in a breach of secondary containment.

Furthermore, this analysis has shown without credit for HVAC [heating,

ventilation, and air conditioning] operation, there would also be a

short term increase in the reactor building ambient temperature.

However, when compared to the UFSAR [Updated Final Safety Analysis

Report] LOCA [loss-of-coolant accident] analysis and the UFSAR main

steam line break analysis, it can be concluded that the consequences

relative to offsite doses, reactor building pressures and temperatures

are bounded by previously analyzed accidents. This change will require

that secondary containment be OPERABLE and capable of handling airborne

radioactivity from steam leaks that could occur during the performance

of hydrostatic pressure or inservice leak testing. Requiring secondary

containment to be OPERABLE will conservatively ensure that, in the

absence of a pipe break, potential airborne radiation from steam leaks

will be filtered through the Standby Gas Treatment System, thereby

minimizing radiation releases to the environment. Leaks to secondary

containment would typically be detected by leakage inspections before

significant inventory loss occurred. This is an integral part of the

hydrostatic pressure and inservice leak testing program. In addition,

there is no mechanism to impart additional fission products into the

reactor coolant. Since the hydrostatic pressure test is performed after

refueling, few noncondensible gases remain in the reactor coolant. In

the proposed condition, the stored energy in the reactor core will be

the same as that at 212 deg.F. This stored energy is sufficiently low

such that even with the loss of inventory following a recirculation

line break, the core coverage could be maintained and the fuel would

not exceed its peak clad temperature limit. Therefore, no significant

release of fission products would occur. Therefore, this change will

not involve a significant increase in the probability or consequences

of an accident previously evaluated.

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 changes to plant

structures, systems, or components (no new or different type of

equipment will be installed and no equipment will be removed). The

change will not alter assumptions made in the safety analyses.

Therefore, the change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

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

safety?

The proposed change allows RCS hydrostatic pressure and leakage

testing to be performed with average reactor coolant temperature in

excess of 212 deg. F and not consider the plant to be in MODE 3

provided certain conditions are met. Secondary containment will be

required to be maintained during the test and all required systems with

the reactor in MODE 4 [cold shutdown] will be OPERABLE in accordance

with the Technical Specifications. Since the hydrostatic or leak tests

are performed water solid, at low decay heat values, and near MODE 4

conditions, the stored energy in the reactor core will be very low.

Under these conditions, the potential for failed fuel and a subsequent

increase in coolant activity is minimized. The reactor pressure vessel

would rapidly depressurize in the event of a large primary system leak

and the low pressure injection systems normally OPERABLE in MODE 4

would be adequate to keep the core flooded. This would ensure that the

fuel would not be uncovered and would not exceed the 2200 deg. F peak

clad temperature limit. Moreover, requiring secondary containment,

including isolation capability, to be OPERABLE will assure that

potential airborne radiation from small leaks can be filtered through

the Standby Gas Treatment System. This will ensure that doses remain

within the limits of 10 CFR 100 guidelines. The potential doses from

any leak or pipe break during the test are bounded by design basis

accident doses presented in the UFSAR. Small system leaks would be

detected by inspections before significant inventory loss has occurred.

In addition, the change provides the benefit of avoiding

depressurization and repressurization of the reactor pressure vessel

during system hydrostatic or

[[Page 6980]]

leakage pressure tests because of the lack of sufficient margin to the

MODE 4/MODE 3 reactor coolant temperature transition limit. Therefore,

the proposed change does not involve a significant reduction in a

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of amendment request: November 1, 1996.

Description of amendment request: The proposed change adds explicit

exceptions to 10 CFR 50 Appendix J in the primary containment leakage

testing program which were previously approved by the Nuclear

Regulatory Commission for the Brunswick Steam Electric Plant Units 1

and 2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change involves reformatting, renumbering, and

rewording the existing Technical Specifications. The reformatting,

renumbering, and rewording process involves no technical changes to the

existing Technical Specifications. As such, this change is

administrative in nature and does not impact initiators of analyzed

events or assumed mitigation of accident or transient events.

Therefore, this change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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 a physical alteration of the

plant (no new or different type of equipment will be installed) or

changes in methods governing normal plant operation. The proposed

change will not impose any new or eliminate any old requirements. Thus,

this change does not create the possibility of a new or different kind

of accident from any accident previously evaluated.

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

safety?

The proposed change will not reduce a margin of safety because it

has no impact on any safety analyses assumptions. This change is

administrative in nature. 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: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: William M. Dean.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County, North Carolina

Date of amendment request: November 1, 1996.

Description of amendment request: The proposed change would change

the requirement of the Rod Block Monitor (RBM) to be Operable when

Thermal Power is greater than or equal to 29% of Rated Thermal Power

and less than 90% of the Rated Thermal Power with the minimum critical

power ratio (MCPR) less than 1.70.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change provides more stringent requirements for

operation of the facility. These more stringent requirements do not

result in operation that will increase the probability of initiating an

analyzed event and do not alter assumptions relative to mitigation of

an accident or transient event. The more restrictive requirements

continue to ensure process variables, structures, systems, and

components are maintained consistent with the safety analyses and

licensing basis. Therefore, this change does not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

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 a physical alteration of the

plant (no new or different type of equipment will be installed) or

changes in the methods governing normal plant operation. The proposed

change does impose different requirements. However, these changes are

consistent with the assumptions in the safety analyses and licensing

basis. Thus, this change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

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

safety?

The imposition of more restrictive requirements either has no

impact on or increases the margin of plant safety. As provided in the

discussion of the change, each change in this category is by

definition, providing additional restrictions to enhance plant safety.

The change maintains requirements within the safety analyses and

licensing basis. Therefore, this change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

[[Page 6981]]

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of amendment request: November 1, 1996.

Description of amendment request: A Rod Worth Minimizer (RWM)

CHANNEL FUNCTIONAL TEST is currently required to be performed during

both a shutdown and a startup. The amendment request would modify the

test frequency to require that the CHANNEL FUNCTIONAL TEST only be

performed once provided the last test performance occurred within a 92-

day period.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

CTS [Current Technical Specification] 4.1.4.1.1 requires a CHANNEL

FUNCTIONAL TEST to be performed prior to withdrawal of control rods for

the purpose of making the reactor critical and when the RWM is

initiated during a plant shutdown. ITS [Improved TS] Surveillance

Requirements are similar to CTS 4.1.4.1.1 except a test Frequency is

specified (92 days). The proposed change effectively extends a[n] RWM

Surveillance Frequency, i.e., the CHANNEL FUNCTIONAL TEST is not

required to be performed if a startup or shutdown occurs within 92 days

of a previous startup or shutdown. The RWM and associated Surveillance

Requirements are not assumed as initiators of any previously analyzed

accidents. In addition, operating history has shown that the RWM would

be continually reliable during the extended Surveillance interval. The

consequences of an accident are not affected by relaxing the Frequency

of the Surveillance since the consequences of a design basis accident

with the RWM inoperable during a reactor startup or shutdown (due to an

undetected failure) are the same as the consequences of a design basis

accident with the RWM inoperable for the proposed 92 day period.

Additionally, the most common outcome of the performance of a

Surveillance is the successful demonstration that the acceptance

criteria are satisfied. This change does not alter assumptions relative

to the mitigation of an accident or transient event. Therefore, this

change does not significantly increase the probability or consequences

of a previously analyzed accident.

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

kind of accident from any accident previously evaluated?

The change introduces no new mode of plant operation and it does

not involve physical modification to the plant. Therefore, it does not

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

accident previously evaluated.

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

safety?

The proposed change to the Frequency is acceptable since the ITS

Surveillance Frequency is adequate for ensuring the RWM is maintained

OPERABLE.

Operating history has shown that the RWM would be continually

reliable during the extended Surveillance interval. The most common

outcome of the performance of a Surveillance is the successful

demonstration that the acceptance criteria are satisfied. Also, the

proposed change provides a benefit of eliminating unnecessary testing

prior to startup and during a shutdown which reduces wear on the

instruments, thereby increasing overall reliability. As such, this

change does not involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North

Carolina

Date of amendment request: December 16, 1997.

Description of amendment request: The amendment request proposes to

revise the Technical Specifications for the Shearon Harris Nuclear

Plant. Specifically, the amendment request proposes revisions to TS

4.7.1.2.1.a.2.a, Auxiliary Feedwater System Surveillance Requirements,

to change the differential pressure and flow requirements of the steam

turbine-driven Auxiliary Feedwater (AFW) pump to allow testing of the

pump at a lower speed than is currently performed.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously evaluated.

Changing the recirculation flow test parameters at which the

turbine-driven AFW pump is tested will demonstrate pump operability

while allowing the surveillance to be performed at a speed that is less

detrimental to the pump. Appropriate testing will continue to ensure

that the Auxiliary Feedwater System (AFS) is capable of performing its

intended function. The proposed amendment will not introduce any new

equipment or require existing equipment to function different from that

previously evaluated in the Final Safety Analysis Report (FSAR) or TS.

Therefore, the proposed change does not involve a significant increase

in the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated.

Changing the recirculation flow test parameters at which the

turbine-driven AFW pump is tested will demonstrate pump operability

while allowing the surveillance to be performed at a speed that is less

detrimental to the pump. Appropriate testing will continue to ensure

that the AFS is capable of performing its intended function. The

proposed amendment will not introduce any new equipment or require

existing equipment to function different from that previously evaluated

in the Final Safety Analysis Report (FSAR) or TS.

[[Page 6982]]

The proposed amendment will not create any new accident scenarios,

because the change does not introduce any new single failures, adverse

equipment or material interactions, or release paths. Therefore, the

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

kind of accident from any accident previously evaluated.

3. The proposed amendment does not involve a significant reduction

in the margin of safety.

Changing the recirculation flow test parameters at which the

turbine-driven AFW pump is tested will demonstrate pump operability

while allowing the surveillance to be performed at a speed that is less

detrimental to the pump. Appropriate testing will continue to ensure

that the AFS is capable of performing its intended function. Therefore,

the proposed change does not involve a significant reduction in the

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: William M. Dean.

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

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

Date of amendment request: December 12, 1997.

Description of amendment request: The proposed amendments would

modify the bypass logic for Main Steam Line Isolation Valve Isolation

Actuation Instrumentation on Condenser Low Vacuum as stated in

Technical Specification (TS) Tables 3.3.2-1 and 4.3.2.1-1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

The reactor vessel steam dome pressure switches, which are proposed

to be removed from the Main Steam Isolation Valve (MSIV) closure scram

bypass logic and the Condenser Vacuum--Low MSLIV [main steam line

isolation valve] isolation bypass logic cause the above trip functions

to become active when the reactor mode switch is not in the RUN

position and the reactor pressure is greater than 1043 psig. The

setpoints of the reactor vessel steam dome pressure switches are the

same as the reactor vessel steam dome pressure--high scram function.

Also, any pressure transients as a result of MSIV closure when not in

Operational Condition 1, Run mode, are minor due to low steam flow

compared to the same event at rated power. Therefore, the reactor

pressure switches being removed from the bypass logic of the MSIV

closure scram has little or no affect on reactor startup, operation,

shutdown, or analyzed accidents.

The condenser vacuum--low isolation function bypass is interlocked

by the same pressure switches that bypass the MSIV closure scram when

the reactor mode switch is not in the RUN position. In addition to

reactor pressure not high, the bypass of the condenser vacuum--low is

bypassed only if the reactor mode switch is not in the RUN position,

all Turbine Stop Valves (TSVs) are not full open, and the keylock

bypass switches are in BYPASS (one for each channel).

With the reactor pressure interlock removed, the remaining

interlocks assure that the condenser will not be overpressurized in

Operational Conditions 2 and 3. The Reactor mode switch interlock

limits reactor thermal power to less than about 12 percent in

Operational Condition 2 (Control Rod withdrawal block on APRM [average

power range monitor] High setpoint in Operational Conditions 2 and 5)

and to much less than 1 percent power when all control rods are fully

inserted in Operational Condition 3 after initial thermal power decay

due to decay heat following reactor shutdown. The Turbine bypass valves

can not be opened with condenser vacuum low (approximately the same as

the isolation setpoint, but different instrumentation). The TSVs remain

closed with condenser vacuum low due to a turbine trip on low condenser

vacuum. Therefore, the remaining bypass interlocks assure that the

isolation of the main steam lines will occur when needed to prevent

overpressurization of the main condenser when vacuum is low or gone.

The change to the position information in the TS Table notes for

the TSV bypass interlock corrects misinformation in the TS. The design

has always used contacts from the auxiliary relays associated with the

``not-full-open'' limit switches for the MSIV closure scram. Therefore,

the setpoints are the same as the MSIV closure scram in TS 2.2.1. The

setpoint in the notes * are made approximate to avoid conflict with the

RPS [reactor protection system] setpoints, which are controlling. Also,

[sic] surveillances for the RPS function for TSV closure scram will

continue to be performed per TS 4.3.1 at the frequencies specified in

TS Table 4.3.1.1-1.

The setpoint for the TSV interlock is not a critical parameter for

the isolation bypass interlock, since the normal position of the TSVs

with low condenser vacuum is fully closed. Therefore, the use of an

approximate value is sufficient, since the actual setpoints and

surveillances are controlled by other specifications.

The reactor pressure switches being removed from the above bypass

circuits are not used for the mitigation of any analyzed accidents or

transients and may actually [decrease] the probability of a scram or

isolation in Startup mode due to the potential for misoperation. Also,

the correction to the TSV position in the bypass notes is more

consistent with the actual setpoints, which are controlled by the

Limiting Safety System Settings for RPS trip function due to TSV

closure.

The rewording of Note * in TS Table 4.3.2.1-1 to be more like Note

* in TS Table 3.3.2-1 helps avoid confusion due to wording differences

and is an administrative type change.

Therefore, there is no significant increase in the probability or

consequences of an accident previously evaluated.

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

from any accident previously evaluated because:

The removal of the reactor pressure switches from the bypass logic

for the MSIV closure scram function and the condenser vacuum--low MSLIV

isolation function with a setpoint equal to the reactor pressure scram

setpoint is not a significant change and does not alter the reactor

modes in which the trips are or can be bypassed. When not in RUN mode,

energy levels are low compared to events that could occur at rated

power levels. These pressure switches only slightly change the bypass

logic and do not affect the scram and isolation circuitry such that a

new or different kind of accident would occur.

The correction of the TSV position interlock for the bypass

function for the condenser vacuum--low MSLIV isolation is not a

physical change to the

[[Page 6983]]

plant, so no failure modes are affected or created.

The rewording of Note * in TS Table 4.3.2.1-1 to be more like Note

* in TS Table 3.3.2-1 helps avoid confusion due to wording differences

and is an administrative type change.

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

is not created.

(3) Involve a significant reduction in the margin of safety

because:

The removal of the rector pressure switches from the bypass logic

of the MSIV closure scram function and the bypass logic from the

condenser vacuum--low MSLIV isolation function does not reduce the

margin of safety, because the setpoints were not established from

analyses that have been performed. The setpoints were set at the value

of the reactor scram on high reactor pressure as a convenient setpoint

out of the way of normal plant operation, rather than initially

removing the bypass interlock.

Also, the high reactor pressure scram is required to be operable in

Operational Conditions 1, 2, and 3, and has no installed means of

bypass, so the removal of the MSIV closure scram in Operational

Conditions other than mode 1, Run mode becoming active due to high

reactor pressure does not reduce the margin for reactor pressurization

events.

The remaining bypass interlocks, associated with TSV position for

the bypass of the condenser vacuum--low MSLIV isolation, assure that

the main condenser will be protected from overpressurization events

with low condenser vacuum. The TSVs are closed due to a main turbine

trip with low condenser vacuum, so if the TSVs were to fail open, the

MSLIV will occur in Operational Conditions 2 and 3 when required. The

removal the reactor pressure bypass interlock and the correction to the

TSV position will not be a significant reduction in the margin of

safety.

The rewording of Note * in TS Table 4.3.2.1-1 to be more like Note

* in TS Table 3.3.2-1 helps avoid confusion due to wording differences

and is an administrative type change.

Therefore, the proposed changes do not involve a significant

reduction in the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: 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, et al., Docket Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: December 11, 1997.

Description of amendment request: The licensee proposed to revise

Table 3.3-4 of the units' Technical Specifications, changing the

Nuclear Service Water System Suction Transfer (from Lake Wylie to the

Standby Nuclear Service Water Pond (SNSWP)) to a higher level of Lake

Wylie. The Nuclear Service Water System is the ultimate heat sink for

various heat loads during normal operation and design basis accidents.

The system also provides makeup water to various systems. Lake Wylie

provides the normal water supply whereas the SNSWP provides an assured

water source should Lake Wylie water becomes unavailable. The transfer

of suction is currently required to occur automatically when Lake

Wylie's levels drops to an elevation of 552.9 feet. The proposed

revision would change this requirement to a more conservative level

about 2.5 feet higher than the current level. This change would correct

previously identified nonconservative aspects of the net positive

suction head (NPSH) calculation for the Nuclear Service Water System

pumps.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92(c). The NRC staff's analysis is

presented below.

1. Will the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

No. The revised suction transfer point would increase reliability

of the Nuclear Service Water System by increasing the NPSH available to

the system. No previously analyzed accidents were initiated by transfer

of the suction source, and the transfer of suction was not a factor in

the consequences of previously analyzed accidents. Therefore, the

proposed change will have no impact on the consequences or

probabilities of any previously evaluated accidents.

2. Will the change create the possibility of a new or difference

kind of accident from any accident previously evaluated?

No. Other than requiring suction be transferred at a higher level

of Lake Wylie, the proposed change would not lead to any hardware or

operating procedure change. Hence, no new equipment failure modes or

accidents from those previously evaluated will be created.

3. Will the change involve a significant reduction in a margin of

safety?

No. Margin of safety is associated with confidence in the design

and operation of the plant. The proposed change to the Technical

Specifications does not involve any change to plant design or

operation. Thus, the margin of safety previously analyzed and evaluated

is maintained.

Based on this analysis, 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: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Attorney for licensee: Mr. Paul R. Newton, Legal Department

(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,

North Carolina.

NRC Project Director: Herbert N. Berkow.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: December 18, 1997; revised on January

26, 1998.

Description of amendment request: The licensee proposed to revise

the units' facility operating licenses (FOL) NPF-35 and NPF-52 to

delete license conditions which have been fulfilled, to update

information to reflect current plant status and regulatory

requirements, and to make other editorial corrections. All the

requested changes are administrative.

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. Will the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

[[Page 6984]]

No. The proposed amendment to the FOL involves administrative

changes only. No actual plant equipment, operating practices, or

accident analyses are affected by this proposed amendment. Therefore,

the proposed amendment has no impact on the possibility (sic) of any

type of accident: new, different, or previously evaluated.

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

kind of accident from any accident previously evaluated?

No. The proposed amendment to the Catawba FOL involves

administrative changes only. No actual plant equipment, operating

practices, or accident analyses are affected by this proposed amendment

and no failure modes not bounded by previously evaluated accidents are

created. Therefore, the proposed amendment has no impact on the

possibility (sic) of any type of accident: new, different, or

previously evaluated.

3. Will the change involve a significant reduction in a margin of

safety?

No. Margin of safety is associated with confidence in the ability

of the fission product barriers (i.e., fuel and fuel cladding, Reactor

Coolant System pressure boundary, and containment structure) to limit

the level of radiation dose to the public. The proposed license

amendment is administrative in nature and only updates the Catawba FOL

to eliminate outdated or completed requirements; therefore, no

reduction in any existing margin of safety is involved.

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: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Attorney for licensee: Mr. Paul R. Newton, Legal Department

(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,

North Carolina.

NRC Project Director: Herbert N. Berkow.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,

Unit No. 1, Pope County, Arkansas

Date of amendment request: December 12, 1997, with supplement dated

August 13, 1997.

Description of amendment request: The proposed amendment

establishes an alternate repair criteria for the segment of steam

generator tubes that are located within the upper tube sheet.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does Not Involve a Significant Increase in the Probability or

Consequences of an Accident Previously Evaluated.

The steam generators are used to remove heat from the reactor

coolant system during normal operation and during accident conditions.

The steam generator tubing forms a substantial portion of the reactor

coolant pressure boundary. A steam generator tube failure is a

violation of the reactor coolant pressure boundary and is a specific

accident analyzed in the ANO-1 Safety Analysis Report.

The purpose of the periodic surveillance performed on the steam

generators in accordance with ANO-1 Technical Specification 4.18 is to

ensure that the structural integrity of this portion of the reactor

coolant system (RCS) will be maintained. The technical specification

plugging limit of 40% of the nominal tube wall thickness requires tubes

to be repaired or removed from service because the tube may become

unserviceable prior to the next inspection. Unserviceable is defined in

the TS as the condition of a tube if it leaks or contains a defect

large enough to affect its structural integrity in the event of an

operating basis earthquake, a loss-of-coolant accident, or a steam line

break.

The proposed technical specification specifies an alternate

plugging limit for upper tubesheet volumetric outer diameter

intergranular attack (ODIGA) indications. Based upon extensive testing

and plant experience, it has been determined that upper tubesheet

volumetric ODIGA flaws with a bobbin voltage indication less than that

specified by the proposed technical specification can remain in service

while maintaining the serviceability of the tube.

From testing performed on simulated flaws within the tubesheet, it

has been shown that the patch IGA indications within the upper

tubesheet, with depths up to 100% through-wall, do not represent

structurally significant flaws which would increase the probability of

a tube failure beyond that currently assumed in the ANO-1 Safety

Analysis Report. The dose consequences of a MSLB accident are analyzed

in the ANO-1 accident analysis. This analysis assumes the unit is

operating with a 1 gpm steam generator tube leak and that the unit has

been operating with 1% defective fuel. Increased leakage during a

postulated MSLB accident resulting from applying the voltage-base

repair criteria to upper tubesheet volumetric ODIGA is not expected.

ODIGA has been present in the ANO-1 steam generators for many years

with no known leakage attributed to this damage mechanism. Because of

its localized nature and morphology, the flaw does not open under

accident conditions. To further support this conclusion, hot leak

testing at the bounding MSLB temperature, pressure, and load was

performed on tubing with representative laboratory generated flaws. The

leak testing was performed on 29 samples with volumetric ODIGA with

bobbin indications of 0.04 to 1.62 volts. None of these flaws showed

signs of leakage as a result of these loads. Additionally, four

specimens created by electrodischarge machining (EDM) with depths up to

approximately 95% through-wall were tested with no leakage detected. It

was, therefore, concluded that volumetric ODIGA flaws with an eddy

current indication up to 1.62 volts will not leak under accident

conditions, and that this is an acceptable threshold value to use to

assume zero accident leakage.

This change allows volumetric ODIGA flaws within the tubesheet,

which are not projected to meet or exceed the 1.62 volt threshold when

considering eddy current uncertainty and an allowance for growth, to

remain in service. Continued operation with these flaws present does

not result in a significant increase in the probability or consequences

of an accident previously evaluated for ANO-1.

Therefore, this change does not involve a significant increase in

the probability or consequences of any accident previously evaluated.

2. Does Not Create the Possibility of a New or Different Kind of

Accident from any Previously Evaluated.

The steam generators are passive components. The intent of the

technical specification surveillance requirements are being met by this

change in that adequate structural and leakage integrity will be

maintained. Additionally, the proposed change does not introduce any

new modes of plant operation.

Therefore, this change does not create the possibility of a new or

different kind of accident from any previously evaluated.

3. Does Not Involve a Significant Reduction in the Margin of

Safety.

The margin of safety is not reduced by the implementation of the

proposed technical specification change allowing

[[Page 6985]]

volumetric ODIGA flaws within the upper tubesheet which meet the

proposed acceptance criteria to remain in service.

Testing of upper tubesheet volumetric ODIGA flaws removed from the

ANO-1 OTSGs during 1R13, showed the flawed tubes to be capable of

withstanding differential pressures of 10,000 psid without the presence

of the tubesheet. Testing of simulated through-wall flaws of up to 0.5

inch in diameter within a tubesheet showed that the tubes always failed

outside of the tubesheet. Thus the structural requirements listed in

the bases of the technical specification are satisfied considering this

change.

Tubes with volumetric ODIGA indications within the tubesheet which

satisfy the acceptance criteria specified in the proposed technical

specification change are not anticipated to leak under accident

conditions. This is due to the small size of the flaws and their

morphology. This premise has been demonstrated through years of actual

plant operation with no known leakage attributable to these flaws, even

considering a plant transient in 1996 which exposed the ``B'' steam

generator to a primary-to-secondary pressure differential of 2100 psid.

The potential for leakage under accident conditions was the focus of

testing performed on representative samples of flawed OTSG tubing.

These tests confirmed for tubesheet flaws, within the bounds of the

proposed technical specification change, that leakage is not expected

under accident conditions. With no increased accident leakage

anticipated as a result of the proposed technical specification change,

the offsite dose consequences from a MSLB accident remain unchanged

from that currently analyzed in the ANO-1 Safety Analysis Report.

Therefore, this change does not involve a significant reduction in

the margin of safety.

In conclusion, based upon the reasoning presented above and the

previous discussion of the amendment request, Entergy Operations has

determined that the requested change does 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: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.

NRC Project Director: John Hannon.

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: November 18, 1996, as supplemented by

letter dated January 21, 1998.

Description of amendment request: The amendment requests a change

to Technical Specification (TS) Surveillance Requirement 4.4.8.3.1.b to

test the Shutdown Cooling System suction line relief valves in

accordance with TS 4.0.5. Editorial changes to 4.4.8.3.1 and

4.4.8.3.1.a. have also been requested.

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. Will operation of the facility in accordance with this proposed

change involve a significant increase in the probability or

consequences of an accident previously evaluated?

No. The proposed change will not affect the assumptions, design

parameters, or results of any accident previously evaluated. The

proposed change does not add or modify any existing equipment. The

proposed change will not diminish the ability of the valves to perform

as required during an accident. The proposed Shutdown Cooling System

suction line relief valves testing schedule will be in accordance with

Section XI of the ASME.

Boiler and Pressure Vessel Code and applicable Addenda as required

by 10 CFR [Part] 50, Section 50.55a(g). This ensures the operational

readiness of the valves. Therefore, the proposed change will not

involve an increase in the probability or consequences of any accident

previously evaluated.

2. Will operation of the facility in accordance with this proposed

change create the possibility of a new or different type of accident

from any accident previously evaluated?

No. The proposed change does not involve modifications to any

existing equipment. The proposed change will not affect the operation

of the plant or the manner in which the plant is operated. No new

failure modes that have not been previously considered will be

introduced. The net effect of the change is to allow the plant staff

the option of reducing the frequency of valve testing to a level that

has been acknowledged as acceptable by the applicable ASME Code.

Therefore, the proposed change will not create the possibility of a new

or different kind of accident from any accident previously evaluated.

3. Will operation of the facility in accordance with this proposed

change involve a significant reduction in a margin of safety?

No. The proposed change does not involve a decrease in the number

or capacity of the valves in the system, nor does it involve a change

in the relief valve setpoints, operability requirements, or limiting

conditions for operation. The margin of safety for the relief valves

is, in part, preserved by compliance with Section XI of the ASME Boiler

and Pressure Vessel Code and applicable Addenda as required by 10 CFR

[Part] 50, Section 50.55a(g). Although the proposed change will allow a

slightly longer testing frequency, the proposed change will continue to

preserve compliance with 10 CFR [Part] 50, Section 50.55a(g).

Therefore, the proposed change will not involve a reduction in a margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of New Orleans

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

Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L

Street N.W., Washington, D.C. 20005-3502.

NRC Project Director: John N. Hannon.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie

Plant, Unit No. 2, St. Lucie County, Florida

Date of amendment request: December 29, 1997.

Description of amendment request: The licensee proposed to modify

specifications for selected cycle-specific reactor physics parameters

so that they refer to the St. Lucie Unit 2 Core Operating Limits Report

(COLR) for limiting values. Minor administrative changes are also

included. The proposed Technical Specification (TS) changes utilized

the guidance provided in Generic Letter 88-16 and are intended to be

consistent with the Standard Technical Specifications for Combustion

Engineering Plants (NUREG-1432, Revision 1).

[[Page 6986]]

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) Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The proposed amendment relocates the calculated values of selected

cycle-specific reactor physics parameter limits from the TS to the

COLR, and includes minor editorial changes which do not alter the

intent of stated requirements. The amendment is administrative in

nature and has no impact on any plant configuration or system

performance relied upon to mitigate the consequences of an accident.

Parameter limits specified in the COLR for this amendment are not

changed from the values presently required by Technical Specifications.

Future changes to the calculated values of such limits may only be made

using NRC approved methodologies, must be consistent with all

applicable safety analysis limits, and are controlled by the 10 CFR

50.59 process. Assumptions used for accident initiators and/or safety

analysis acceptance criteria are not changed by this amendment.

Therefore, operation of the facility in accordance with the proposed

amendment will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

(2) Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different kind

of accident from any accident previously evaluated.

The proposed amendment relocates the calculated values of cycle

specific reactor physics limiting parameters to the COLR and will not

change the physical plant or the modes of operation defined in the

facility license. The changes do not involve the addition of new

equipment or the modification of existing equipment, nor do they alter

the design configuration of St. Lucie plant systems. Therefore,

operation of the facility in accordance with the proposed amendment

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

from any accident previously evaluated.

(3) Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The cycle specific parameter limits being relocated to the COLR by

this amendment have not been changed from the values presently required

by the TS, and a requirement to operate the plant within the bounds of

the limits specified in the COLR is retained in the individual

specifications. Future changes to the calculated values of these limits

by the licensee may only be developed using NRC-approved methodologies,

must remain consistent with all applicable plant safety analysis limits

addressed in the Final Safety Analysis Report (FSAR), and are further

controlled by the 10 CFR 50.59 process. As discussed in Generic Letter

88-16, the administrative controls established for the values of cycle

specific parameters using the guidance of that letter assure

conformance with 10 CFR 50.36. Safety analysis acceptance criteria are

not being altered by this amendment. Therefore, operation of the

facility in accordance with the proposed amendment would 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 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: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,

P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Project Director: Frederick J. Hebdon.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn

County, Iowa

Date of amendment request: October 30, 1996.

Description of amendment request: The proposed amendment, included

as part of the proposed conversion from current Technical

Specifications (TS) to improved TS, would relax the required flowrates

in core spray, low pressure coolant injection (LPCI), and high pressure

coolant injection (HPCI) systems, based on the DAEC loss-of-coolant-

accident (LOCA) analysis, using an NRC-approved code, SAFER/GESTR-LOCA.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change will lower ECCS required flowrates in

accordance with accident analysis assumptions. The ECCS subsystems

affected by this change are not assumed to be initiators of analyzed

events. Therefore, the proposed change does not increase the

probability of any accident. The role of these ECCS subsystems is in

the mitigation of accident consequences. The proposed change decreases

pump flow rate requirements for Core Spray, LPCI and HPCI. The proposed

change does not increase the consequences of an accident because

accident analysis presented in NEDC-31310P, Duane Arnold Energy Center

SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis, uses these reduced

pump flow rates as analysis inputs and demonstrates that peak cladding

temperatures are maintained within regulatory limits. Therefore, this

change will not involve a significant increase in the consequences of

an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change will not involve any physical changes to plant

systems, structures, or components (SSCs), or the manner in which these

SSCs are operated, maintained, modified, tested, or inspected. As

demonstrated in NEDC-31310P, Duane Arnold Energy Center SAFER/GESTR-

LOCA Loss-of-Coolant Accident Analysis, at the reduced flowrates,

adequate ECCS capability will still exist to mitigate the consequences

of accidents. Therefore, this change will not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

safety?

The proposed change does not significantly reduce the margin of

safety because accident analysis presented in NEDC-31310P, Duane Arnold

Energy Center SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis, uses

these reduced pump flow rates as analysis inputs. The accident analysis

demonstrates that with these reduced ECCS pump flow rates, the peak

clad temperature remains below the regulatory limit. Therefore, this

change does not involve a significant reduction in a margin of safety.

[[Page 6987]]

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: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids, Iowa 52401.

Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan,

Lewis, & Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

Acting NRC Project Director: Richard P. Savio.

IES Utilities Inc., Docket No. 50-331 Duane Arnold Energy Center, Linn

County, Iowa

Date of amendment requests: January 9, 1998.

Description of amendment requests: The proposed amendment would

revise the limiting condition for operation for primary containment

isolation valves (PCIVs). The revision would allow 72 hours to isolate

a failed valve associated with a closed system.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment will not involve a significant increase

in the probability or consequences of an accident previously evaluated.

This change extends the time to isolate single PCIV penetrations

from 4 hours to 72 hours. The time allowed to isolate the penetration

is not assumed to be an initiator of any analyzed event. The 72 hour

period provides the necessary time to perform repairs on a failed

containment isolation valve when relying on an intact closed system.

Use of a closed system for isolation is directly equivalent to

isolating a failed containment isolation valve by use of a single

valve. The closed systems are subject to a Type A containment leakage

test, are missile protected, and are seismic Category 1 piping.

Allowing an additional 68 hours to isolate these penetrations will not

significantly increase the consequences of an accident since the intact

closed system provides adequate isolation. Also, the consequences of an

event occurring during the proposed 72 hour period are the same as

those during the current 4 hour period. The 72 hour period is

consistent with NRC-approved Traveler TSTF-30, Revision 2. Therefore,

this change does not involve a significant increase in the probability

or consequences of an accident previously evaluated.

2. The proposed amendment will not create the possibility of a new

or different kind of accident from any accident previously evaluated.

This change extends the time allowed to isolate single PCIV

penetrations from 4 hours to 72 hours. The additional 68 hours that the

penetrations are not isolated will not create the possibility of a new

or different kind of accident. Use of a closed system for isolation is

directly equivalent to isolating a failed containment isolation valve

by use of a single valve. The closed systems are subject to a Type A

containment leakage test, are missile protected, and are seismic

Category 1 piping. This change will not physically alter the plant (no

new or different type of equipment will be installed). The change in

allowed out-of-service-time is consistent with current safety analysis

assumptions. Therefore, this change does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed amendment does not involve a significant reduction

in a margin of safety.

This change extends the time allowed to isolate single PCIV

penetrations from 4 hours to 72 hours. During the additional time

allowed, a limiting event would still be assumed to be within the

bounds of the safety analysis assuming no single active failure. The 72

hour period is consistent with NRC-approved Traveler TSTF-30, Revision

2. Use of a closed system for isolation is directly equivalent to

isolating a failed containment isolation valve by use of a single

valve. Therefore, this change does not involve a significant reduction

in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Cedar Rapids Public Library,

00 First Street, SE., Cedar Rapids, Iowa 52401.

Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &

Brockius, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Acting Project Director: Richard P. Savio.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,

Unit No. 1, Washington County, Nebraska

Date of amendment request: December 11, 1997.

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TS) to add a new Limiting

Condition for Operation (LCO) for an inoperable engineering safety

features (ESF) logic subsystem.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

Omaha Public Power District (OPPD) proposes to incorporate a new

Limiting Condition for Operation (LCO) into Specification 2.15 which

will apply to an engineered safety features (ESF) logic subsystem when

the minimum operable channels or minimum degree of redundancy

requirements listed in Tables 2-3 and 2-4 are not met. The LCO proposes

an allowed outage time (AOT) of 48 hours to restore sufficient channels

to operability so as to exceed minimum requirements, or the plant must

be placed in hot shutdown within the following 12 hours.

The ESF logic system is a Class 1 protection system designed to

satisfy the criteria of IEEE 279, August 1968. Two functionally

redundant ESF logic subsystems ``A'' and ``B'' are provided to ensure

high reliability and effective in-service testing. These logic

subsystems are designed for individual reliability and maximum

attainable mutual independence both physically and electrically. Either

ESF logic subsystem acting alone can automatically actuate ESF

equipment and essential supporting systems.

The design of the ESF logic system is not being altered by this

change. The change allows a reasonable time to contact trained

personnel and adequately troubleshoot, perform and test repairs on an

inoperable ESF logic subsystem. The proposed AOT ensures that repairs

are thoroughly planned and accomplished without undue haste. In this

situation, the opposite ESF logic subsystem is operable as verified

through surveillance testing and capable of providing both automatic

and manual ESF equipment actuation.

The proposed AOT is similar to that of LCO 3.3.5, ``Engineered

Safety Features Actuation System (ESFAS)

[[Page 6988]]

Logic and Manual Trip (Analog),'' of Combustion Engineering Owners

Group (CEOG) Standard Technical Specification (STS), Rev. 1, dated

April 7, 1995.

Additional administrative revisions are proposed to either support

the new LCO (e.g., footnotes in Tables 2-3 & 2-4) or clarify existing

information. Therefore, OPPD concludes that the proposed LCO and

administrative revisions do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

different kind of accident from any accident previously evaluated.

There will be no physical alterations to the plant configuration,

changes to setpoint values, or changes to the application of setpoints

or limits because of these proposed changes. No changes in operating

modes are proposed. The proposed LCO provides a reasonable AOT to

troubleshoot, repair, and test an inoperable ESF logic subsystem. The

remaining ESF logic subsystem is still operable and capable of both

automatic and manual ESF equipment actuation. The remaining changes are

administrative in nature and thus none of the proposed changes 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.

The proposed LCO provides a reasonable AOT to troubleshoot, repair,

and test an inoperable ESF logic subsystem. The remaining ESF logic

subsystem is still operable as verified by surveillance testing and

capable of both automatic and manual ESF equipment actuation. With an

inoperable ESF logic subsystem, the ESF logic system would not be

single failure proof for a brief period of time. However, it is OPPD's

position that making repairs while the plant is at power and stable is

preferable to imposing a transient (manual shutdown) on the plant at a

time when the ESF logic system is no longer single failure proof.

Therefore, OPPD concludes that the proposed LCO and supporting

administrative changes do not result in a significant reduction in a

margin of safety.

Based on the above considerations, it is OPPD's position that this

proposed amendment does not involve significant hazards considerations

as defined by 10 CFR 50.92 and the proposed changes will not result in

a condition which significantly alters the impact of the Station on the

environment. Thus, the proposed changes meet the eligibility criteria

for categorical exclusion set forth in 10 CFR 51.22(c)(9) and pursuant

to 10 CFR 51.22(b) no environmental assessment need be prepared.

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: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102.

Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005-3502.

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: January 12, 1998.

Description of amendment request: The Philadelphia Electric Company

submitted a Technical Specifications (TS) Change Request, requesting an

amendment to the TS (Appendix A) of Operating License No. NPF-39 for

Limerick Generating Station (LGS), Unit 1. This proposed change will

revise TS Table 4.4.6.1.3-1 to change the withdrawal schedule for the

first capsule to be withdrawn from 10 Effective Full Power Years (EFPY)

to 15 EFPY.

A revision to TS Surveillance Requirement 4.4.6.1.4 is also

proposed. This revision will remove the references to flux wire removal

and analysis that was originally required following the first cycle of

operation. The referenced flux wires were never located following the

first cycle of operation. This TS Surveillance Requirement will be

changed to refer to the flux wires that are located within the

surveillance capsules, which will be removed and analyzed in accordance

with the surveillance capsule removal schedule located in TS Table

4.4.6.1.3-1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed Technical Specifications (TS) changes do not

involve a significant increase in the probability or consequences of an

accident previously evaluated.

The proposed changes do not increase the probability of occurrence

of an accident previously evaluated in the safety analysis report and

do not affect any accident initiators as described in the SAR [Safety

Analysis Report]. The changes revise the withdrawal schedule for the

reactor vessel material surveillance capsules from 10 Effective Full

Power Years (EFPY) to 15 EFPY. The capsules are not an initiator of any

previously analyzed accident nor does the withdrawal schedule of the

surveillance capsule affect the probability or consequences of any

previously analyzed accident.

These changes will not affect the Pressure-Temperature (P-T) limits

as given in LGS Technical Specification (TS) Figure 3.4.6.1-1 and UFSAR

[Updated Final Safety Analysis Report] Figure 5.3-4. P-T limits are

imposed on the reactor coolant system to ensure that adequate safety

margins exist during normal operation, anticipated operational

occurrences, and system hydrostatic tests. The P-T limits are related

to the RTNDT [reference temperature], as described in ASME

Section III, Appendix G. Changes in the fracture toughness properties

of reactor pressure vessel (RPV) beltline materials, resulting from

neutron irradiation and the thermal environment, are monitored by a

surveillance program in compliance with the requirements of 10 CFR 50

Appendix H. The effect of neutron fluence on the shift in the

RTNDT is predicted by methods given in Regulatory Guide

1.99, Rev. 2.

As detailed in Attachment 3 [of the licensee's application dated

January 12, 1998], for LGS Unit 1, the combination of low expected

RTNDT shift for the plate material due to low predicted

fluence and excellent material chemistry, Supplemental Surveillance

Program (SSP) data on similar material, and the inherent margin in the

P-T curve calculations--with the withdrawal schedule of the first

surveillance capsule modified from 10 EFPY to 15 EFPY--will result in a

more credible set of surveillance data while ensuring the continued

safe operation of LGS Unit 1.

LGS's current P-T limits were established based on adjusted

reference temperatures developed in accordance with the procedures

prescribed in Regulatory Guide 1.99, Rev. 2, Regulatory Position 1,

``Surveillance Data Not Available.'' Calculation of adjusted reference

temperature by these procedures includes a conservative base fluence

estimate, power rerate adjustment of a 110% fluence multiplier from

startup--instead of a 105% fluence

[[Page 6989]]

multiplier since 1R06 [Unit 1 refueing outage 6], and a margin term to

ensure conservative, upper-bound values are used for the calculation of

the P-T limits. Revision of the first capsule withdrawal schedule will

not affect the P-T limits because the capsule constitutes one set of

credible surveillance data. The curves will continue to be established

in accordance with Regulatory Position 1 procedures.

As per Regulatory Guide 1.99, Radiation Embrittlement of Reactor

Vessel Materials, Revision 2, Regulatory Position 2, ``Surveillance

Data Available,'' the collection of two or more sets of credible

surveillance data is necessary to empirically calculate the adjusted

reference temperature (ART). Each surveillance capsule constitutes one

set of credible surveillance data. This calculated ART can be used to

revise the Pressure-Temperature (P-T) curves (Technical Specification

Figure 3.4.6.1-1). Without two or more sets of credible data, the ART

must be calculated and the P-T curves revised, based upon the

calculational methodologies as provided in the Regulatory Guide 1.99,

Rev. 2, Regulatory Position 1, ``Surveillance Data Not Available.''

These methodologies use plant specific chemistry and fluence values to

determine a calculated shift in RTNDT. A ``margin'' term is

then added to obtain conservative, upper-bound values of adjusted

reference temperature.

The existing LGS Unit 1 P-T curves are currently valid up to 12

EFPY. With first capsule removal at either 10 or 15 EFPY, the existing

P-T curves will require a revision prior to reaching 12 EFPY based upon

the calculational methodologies as contained in the Regulatory Guide

1.99, Rev. 2, Regulatory Position 1, ``Surveillance Data Not

Available.'' Therefore, the revision to the first capsule withdrawal

schedule results in no impact to the calculational methodologies that

will be used for the P-T curve revision that will be necessary to

extend the curves beyond 12 EFPY.

The fluence data as determined from the surveillance capsule flux

wires at 15 EFPY will provide an accurate indication of neutron

fluence. In accordance with Regulatory Guide 1.99, Rev. 2, Regulatory

Position 1 methodology, data from these flux wires will permit an

adjustment of TS Figure 3.4.6.1-1 in accordance with TS surveillance

requirement 4.4.6.1.3, if required, and will meet the requirements of

10 CFR 50 Appendix H and ASTM E-185.

These changes will not affect any plant safety limits or limiting

conditions of operation. The proposed changes will not affect reactor

pressure vessel performance as they do not involve any physical

changes, and LGS P-T limits will remain conservative in accordance with

Reg. Guide 1.99, Rev. 2 requirements. The proposed changes will not

cause the RPV or interfacing systems to be operated outside of their

design or testing limits.

The proposed changes do not increase the consequences of a

malfunction of equipment important to safety previously evaluated in

the SAR. The proposed changes do not involve any physical changes to

equipment important to safety. The potential for RPV failure will be

adequately assessed by the proposed withdrawal schedule. In addition,

the results from the SSP will provide industry data that bounds the

materials used in the LGS Unit 1 reactor pressure vessel until the data

from the first LGS Unit 1 capsule is available. The proposed changes

provide the same level of confidence in the integrity of the vessel.

Therefore, the proposed TS changes do not involve an increase in

the probability or consequences of an accident previously evaluated.

2. The proposed TS changes do 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 different

type of accident than any previously evaluated in the SAR. The proposed

changes will revise the withdrawal schedule for the first reactor

pressure vessel (RPV) material surveillance capsule from 10 Effective

Full Power Years (EFPY) to 15 EFPY. These proposed changes do not

involve a physical modification of the design of plant structures,

systems or components. The proposed changes will not impact the manner

in which the plant is operated, as plant operating and testing

procedures will not be affected by the changes. No new accident types

or failure modes will be introduced as a result of the proposed

changes.

LGS's current Pressure-Temperature (P-T) limits were established

based on adjusted reference temperatures developed in accordance with

the procedures prescribed in Regulatory Guide 1.99, Rev. 2, Regulatory

Position 1, ``Surveillance Data Not Available.'' Calculation of

adjusted reference temperature by these procedures includes a

conservative base fluence estimate, power rerate adjustment of a 110%

fluence multiplier from startup--instead of a 105% fluence multiplier

since 1R06, and a margin term to ensure conservative, upper-bound

values are used for the calculation of the P-T limits. Revision of the

first capsule withdrawal schedule will not affect the P-T limits

because the capsule constitutes one set of credible surveillance data.

The curves will continue to be established in accordance with

Regulatory Position 1 procedures.

The existing LGS Unit 1 P-T curves are currently valid up to 12

EFPY. With first capsule removal at either 10 or 15 EFPY, the existing

P-T curves will require a revision, prior to reaching 12 EFPY, based

upon the calculational methodologies as contained in the Regulatory

Guide 1.99, Rev. 2, Regulatory Position 1, ``Surveillance Data Not

Available.''

Therefore, the Technical Specification (TS) revision to the first

capsule withdrawal schedule results in no impact to the calculational

methodologies that will be used for the P-T curve revision that will be

necessary to extend the curves beyond 12 EFPY.

The fluence data as determined from the surveillance capsule flux

wires at 15 EFPY will provide an accurate indication of neutron

fluence. In accordance with Regulatory Guide 1.99, Rev. 2, Regulatory

Position 1 methodology, data from these flux wires will permit an

adjustment of TS Figure 3.4.6.1-1 in accordance with TS Surveillance

Requirement 4.4.6.1.3, if required, and will meet the requirements of

10 CFR 50 Appendix H and ASTM E-185.

The potential for reactor pressure vessel (RPV) failure will

continue to be adequately assessed by the proposed withdrawal schedule.

As detailed in Attachment 3, the combination of the low expected shift

for the plate material, SSP data on similar material, and the inherent

margin in the P-T curve calculations will result in a credible set of

surveillance data, while ensuring the continued safe operation of LGS

Unit 1. The proposed changes provide the same level of confidence in

the integrity of the RPV.

Therefore, the proposed TS changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed TS changes do not involve a significant reduction

in a margin of safety.

The proposed changes to the Technical Specifications (TS) do not

reduce the margin of safety as defined in the Bases for any TS. The

proposed changes will not affect any safety limits, limiting safety

system settings, or limiting conditions of operation. The proposed

changes do not represent a change in initial conditions, system

response time, or in any other parameter

[[Page 6990]]

affecting the accident analyses supporting the Bases of any TS. The

proposed changes do not involve revision of the P-T limits but rather a

revision of the withdrawal schedule for the first surveillance capsule.

The current P-T limits were established based on the adjusted reference

temperatures for vessel beltline materials calculated in accordance

with Regulatory Position 1 of Reg. Guide 1.99, Rev. 2. P-T limits will

continue to be revised as necessary for changes in adjusted reference

temperature due to changes in fluence according to Regulatory Position

1 until two or more credible surveillance data sets become available.

When two or more credible surveillance data sets become available, P-T

limits will be revised as prescribed by Regulatory Position 2 of Reg.

Guide 1.99, Rev. 2 or other NRC approved guidance.

The current P-T limit curves are inherently conservative and

provide sufficient margin to ensure the integrity of the reactor

pressure vessel. The proposed changes do not adversely affect these

curves. The fluence data as determined from the surveillance capsule

flux wires at 15 EFPY will provide an accurate indication of neutron

fluence.

In accordance with Regulatory Guide 1.99, Rev. 2, Regulatory

Position 1 methodology, data from these flux wires will permit an

adjustment of TS Figure 3.4.6.1-1 in accordance with TS Surveillance

Requirement 4.4.6.1.3, if required, and will meet the requirements of

10 CFR 50 Appendix H and ASTM E-185.

Therefore, the proposed TS changes do not involve a reduction in a

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,

Limerick Generating Station, Units 1 and 2, Montgomery County,

Pennsylvania

Date of amendment request: September 2, 1997.

Description of amendment request: This proposed Technical

Specification (TS) Change Request revises TS Sections 4.0.5, and Bases

Sections B 4.0.5 and B 3/4.4.8, for Limerick Generating Station (LGS),

Units 1 and 2, pertaining to the surveillance requirement associated

with Inservice Inspection (ISI) and Inservice Testing (IST) activities

for American Society of Mechanical Engineers (ASME) Boiler and Pressure

Vessel (B&PV) Code, Class 1, 2, and 3 components.

The existing wording in TS Section 4.0.5, and Bases Sections B

4.0.5 and B 3/4.4.8, stipulates that ISI and IST surveillance

activities for ASME Code Class 1, 2, and 3 components be conducted in

accordance with the requirements of Section XI of the ASME Code as

required by 10 CFR 50.55a(g). The proposed changes will revise the

applicable TS sections to only make reference to 10 CFR 50.55a, since

the current regulations have separated the specific requirements for

ISI and IST into sections 50.55a(g) and 50.55a(f), respectively.

The existing wording of TS Section 4.0.5, and Bases Sections B

4.0.5 and B 3/4.4.8, also requires that ISI and IST surveillance

activities be conducted in accordance with the requirements of Section

XI of the ASME Boiler and Pressure Vessel Code, except where specific

written relief has been granted by the NRC. This wording precludes the

immediate implementation of alternative testing in the event that a

Code required inspection has been identified as clearly impractical.

The proposed TS changes will revise the applicable TS sections to

eliminate the requirement that written relief be obtained prior to

implementation of alternative testing during the initial 120-month

inspection interval, and the initial 12 months of subsequent intervals

in cases where the Code required inspections have been found to be

clearly impractical. NUREG-1482, ``Guidelines for Inservice Testing at

Nuclear Power Plants,'' discusses impracticality as being a situation

where a test cannot be performed due to limitations in design (which

includes prohibitive dose rates), construction, or system

configuration.

Furthermore, TS Section 4.0.5b. currently discusses the required

frequency of ISI and IST surveillance activities required by the ASME

Code. The existing TS address testing frequencies of up to one (1)

year. In some cases, the ASME Code requires that testing be performed

on a two (2) year frequency. The proposed TS changes will also revise

the TS to include a reference for tests that are conducted on a

biennial frequency. Inclusion of this reference will permit the

application of TS 4.0.2 criteria for ISI and IST surveillance

activities. This will permit a 25 percent time extension to be applied

to the surveillance frequency, if necessary, in order to allow for

consideration of plant operating conditions when scheduling ISI and IST

surveillance tests.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed Technical Specifications (TS) changes do not

involve a significant increase in the probability or consequences of an

accident previously evaluated.

The proposed TS changes are administrative in nature and do not

make physical modifications or changes to the plant structures,

systems, or components (SSC). Plant SSC will continue to function as

designed. The proposed TS changes will not alter equipment operational

practices or procedures.

In the event that an ASME Section XI Code required inspection or

test is found to be impractical due to unforeseen conditions, written

relief would still be requested from the NRC in accordance with

established procedures. No code required inspection will be eliminated

from the ISI or IST Programs until written approval has been granted by

the NRC as required [by] 10CFR50.55a. It is anticipated that the only

time this provision would be utilized would be in the event that an

inspection or test is discovered to be impossible or impractical to

perform due to unforeseen or unexpected high radiation conditions, or

physical limitations. This change will also clarify the applicability

of surveillance intervals to biennial tests or examinations.

The proposed TS changes will remove the inconsistencies between the

LGS TS and the requirements of 10CFR50.55a, and will also ensure that

the implementation of the LGS ISI and IST Programs are consistent with

current NRC guidance as specified in NUREG-1482 and NUREG-1433,

Revision 1.

Therefore, the proposed TS changes do not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

2. The proposed TS changes do not create the possibility of a new

or different kind of accident from any accident previously evaluated.

[[Page 6991]]

The proposed changes apply to the administrative requirements for

testing of plant systems. No physical modifications to systems or

components are involved. No new failure modes which could cause or

contribute to the cause of an accident are being introduced.

The proposed TS changes will remove the inconsistencies between the

LGS TS and the requirements of 10CFR50.55a, and will also ensure that

the implementation of the LGS ISI and IST Programs are consistent with

current NRC guidance as specified in NUREG-1482 and NUREG-1433,

Revision 1.

Therefore, the proposed TS changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed TS changes do not involve a significant reduction

in a margin of safety.

No physical plant modifications or operational procedure changes

are being made as a result of the proposed TS changes. The proposed TS

changes apply to the ISI and IST Programs' surveillance requirements

and do not modify the scope or frequency of these Programs as required

by 10 CFR 50.55a. The proposed TS changes will eliminate

inconsistencies between current TS wording and the requirements

specified in 10CFR50.55a. In addition, the proposed changes are

consistent with the guidance stipulated in NUREG-1482 and NUREG-1433,

Revision 1. No physical plant modifications or operational procedure

changes are being introduced as a result of this proposed TS Change.

Therefore, the proposed TS changes do not involve a reduction in a

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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.

Power Authority of the State of New York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: October 8, 1997.

Description of amendment request: This amendment proposes revisions

to the actions to be taken in the event multiple control rods are

inoperable.

Basis for proposed no significant hazards conside

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