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

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URL: https://www.frixlaw.com/law-library/documents/fr%3A98-3269

## Record

- **Collection:** Federal Register
- **Document type:** Notice
- **Published:** February 11, 1998
- **Citation:** 63 FR 6968

## Text

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

[[Page 6970]]

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),

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A98-3269. Public record. Not legal advice.