98-3269. Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

[Federal Register Volume 63, Number 28 (Wednesday, February 11, 1998)]
[Notices]
[Pages 6968-7004]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-3269]


-----------------------------------------------------------------------

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), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed amendment relocates the calculated values of selected 
cycle-specific reactor physics parameter limits from the TS to the 
COLR, and includes minor editorial changes which do not alter the 
intent of stated requirements. The amendment is administrative in 
nature and has no impact on any plant configuration or system 
performance relied upon to mitigate the consequences of an accident. 
Parameter limits specified in the COLR for this amendment are not 
changed from the values presently required by Technical Specifications. 
Future changes to the calculated values of such limits may only be made 
using NRC approved methodologies, must be consistent with all 
applicable safety analysis limits, and are controlled by the 10 CFR 
50.59 process. Assumptions used for accident initiators and/or safety 
analysis acceptance criteria are not changed by this amendment. 
Therefore, operation of the facility in accordance with the proposed 
amendment will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed amendment relocates the calculated values of cycle 
specific reactor physics limiting parameters to the COLR and will not 
change the physical plant or the modes of operation defined in the 
facility license. The changes do not involve the addition of new 
equipment or the modification of existing equipment, nor do they alter 
the design configuration of St. Lucie plant systems. Therefore, 
operation of the facility in accordance with the proposed amendment 
would not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The cycle specific parameter limits being relocated to the COLR by 
this amendment have not been changed from the values presently required 
by the TS, and a requirement to operate the plant within the bounds of 
the limits specified in the COLR is retained in the individual 
specifications. Future changes to the calculated values of these limits 
by the licensee may only be developed using NRC-approved methodologies, 
must remain consistent with all applicable plant safety analysis limits 
addressed in the Final Safety Analysis Report (FSAR), and are further 
controlled by the 10 CFR 50.59 process. As discussed in Generic Letter 
88-16, the administrative controls established for the values of cycle 
specific parameters using the guidance of that letter assure 
conformance with 10 CFR 50.36. Safety analysis acceptance criteria are 
not being altered by this amendment. Therefore, operation of the 
facility in accordance with the proposed amendment would not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Project Director: Frederick J. Hebdon.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa

    Date of amendment request: October 30, 1996.
    Description of amendment request: The proposed amendment, included 
as part of the proposed conversion from current Technical 
Specifications (TS) to improved TS, would relax the required flowrates 
in core spray, low pressure coolant injection (LPCI), and high pressure 
coolant injection (HPCI) systems, based on the DAEC loss-of-coolant-
accident (LOCA) analysis, using an NRC-approved code, SAFER/GESTR-LOCA.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change will lower ECCS required flowrates in 
accordance with accident analysis assumptions. The ECCS subsystems 
affected by this change are not assumed to be initiators of analyzed 
events. Therefore, the proposed change does not increase the 
probability of any accident. The role of these ECCS subsystems is in 
the mitigation of accident consequences. The proposed change decreases 
pump flow rate requirements for Core Spray, LPCI and HPCI. The proposed 
change does not increase the consequences of an accident because 
accident analysis presented in NEDC-31310P, Duane Arnold Energy Center 
SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis, uses these reduced 
pump flow rates as analysis inputs and demonstrates that peak cladding 
temperatures are maintained within regulatory limits. Therefore, this 
change will not involve a significant increase in the consequences of 
an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change will not involve any physical changes to plant 
systems, structures, or components (SSCs), or the manner in which these 
SSCs are operated, maintained, modified, tested, or inspected. As 
demonstrated in NEDC-31310P, Duane Arnold Energy Center SAFER/GESTR-
LOCA Loss-of-Coolant Accident Analysis, at the reduced flowrates, 
adequate ECCS capability will still exist to mitigate the consequences 
of accidents. Therefore, this change will not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does this change involve a significant reduction in a margin of 
safety?
    The proposed change does not significantly reduce the margin of 
safety because accident analysis presented in NEDC-31310P, Duane Arnold 
Energy Center SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis, uses 
these reduced pump flow rates as analysis inputs. The accident analysis 
demonstrates that with these reduced ECCS pump flow rates, the peak 
clad temperature remains below the regulatory limit. Therefore, this 
change does not involve a significant reduction in a margin of safety.

[[Page 6987]]

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401.
    Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, 
Lewis, & Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    Acting NRC Project Director: Richard P. Savio.

IES Utilities Inc., Docket No. 50-331 Duane Arnold Energy Center, Linn 
County, Iowa

    Date of amendment requests: January 9, 1998.
    Description of amendment requests: The proposed amendment would 
revise the limiting condition for operation for primary containment 
isolation valves (PCIVs). The revision would allow 72 hours to isolate 
a failed valve associated with a closed system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    This change extends the time to isolate single PCIV penetrations 
from 4 hours to 72 hours. The time allowed to isolate the penetration 
is not assumed to be an initiator of any analyzed event. The 72 hour 
period provides the necessary time to perform repairs on a failed 
containment isolation valve when relying on an intact closed system. 
Use of a closed system for isolation is directly equivalent to 
isolating a failed containment isolation valve by use of a single 
valve. The closed systems are subject to a Type A containment leakage 
test, are missile protected, and are seismic Category 1 piping. 
Allowing an additional 68 hours to isolate these penetrations will not 
significantly increase the consequences of an accident since the intact 
closed system provides adequate isolation. Also, the consequences of an 
event occurring during the proposed 72 hour period are the same as 
those during the current 4 hour period. The 72 hour period is 
consistent with NRC-approved Traveler TSTF-30, Revision 2. Therefore, 
this change does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. The proposed amendment will not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    This change extends the time allowed to isolate single PCIV 
penetrations from 4 hours to 72 hours. The additional 68 hours that the 
penetrations are not isolated will not create the possibility of a new 
or different kind of accident. Use of a closed system for isolation is 
directly equivalent to isolating a failed containment isolation valve 
by use of a single valve. The closed systems are subject to a Type A 
containment leakage test, are missile protected, and are seismic 
Category 1 piping. This change will not physically alter the plant (no 
new or different type of equipment will be installed). The change in 
allowed out-of-service-time is consistent with current safety analysis 
assumptions. Therefore, this change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant reduction 
in a margin of safety.
    This change extends the time allowed to isolate single PCIV 
penetrations from 4 hours to 72 hours. During the additional time 
allowed, a limiting event would still be assumed to be within the 
bounds of the safety analysis assuming no single active failure. The 72 
hour period is consistent with NRC-approved Traveler TSTF-30, Revision 
2. Use of a closed system for isolation is directly equivalent to 
isolating a failed containment isolation valve by use of a single 
valve. Therefore, this change does not involve a significant reduction 
in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Cedar Rapids Public Library, 
00 First Street, SE., Cedar Rapids, Iowa 52401.
    Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis & 
Brockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Acting Project Director: Richard P. Savio.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: December 11, 1997.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to add a new Limiting 
Condition for Operation (LCO) for an inoperable engineering safety 
features (ESF) logic subsystem.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Omaha Public Power District (OPPD) proposes to incorporate a new 
Limiting Condition for Operation (LCO) into Specification 2.15 which 
will apply to an engineered safety features (ESF) logic subsystem when 
the minimum operable channels or minimum degree of redundancy 
requirements listed in Tables 2-3 and 2-4 are not met. The LCO proposes 
an allowed outage time (AOT) of 48 hours to restore sufficient channels 
to operability so as to exceed minimum requirements, or the plant must 
be placed in hot shutdown within the following 12 hours.
    The ESF logic system is a Class 1 protection system designed to 
satisfy the criteria of IEEE 279, August 1968. Two functionally 
redundant ESF logic subsystems ``A'' and ``B'' are provided to ensure 
high reliability and effective in-service testing. These logic 
subsystems are designed for individual reliability and maximum 
attainable mutual independence both physically and electrically. Either 
ESF logic subsystem acting alone can automatically actuate ESF 
equipment and essential supporting systems.
    The design of the ESF logic system is not being altered by this 
change. The change allows a reasonable time to contact trained 
personnel and adequately troubleshoot, perform and test repairs on an 
inoperable ESF logic subsystem. The proposed AOT ensures that repairs 
are thoroughly planned and accomplished without undue haste. In this 
situation, the opposite ESF logic subsystem is operable as verified 
through surveillance testing and capable of providing both automatic 
and manual ESF equipment actuation.
    The proposed AOT is similar to that of LCO 3.3.5, ``Engineered 
Safety Features Actuation System (ESFAS)

[[Page 6988]]

Logic and Manual Trip (Analog),'' of Combustion Engineering Owners 
Group (CEOG) Standard Technical Specification (STS), Rev. 1, dated 
April 7, 1995.
    Additional administrative revisions are proposed to either support 
the new LCO (e.g., footnotes in Tables 2-3 & 2-4) or clarify existing 
information. Therefore, OPPD concludes that the proposed LCO and 
administrative revisions do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    There will be no physical alterations to the plant configuration, 
changes to setpoint values, or changes to the application of setpoints 
or limits because of these proposed changes. No changes in operating 
modes are proposed. The proposed LCO provides a reasonable AOT to 
troubleshoot, repair, and test an inoperable ESF logic subsystem. The 
remaining ESF logic subsystem is still operable and capable of both 
automatic and manual ESF equipment actuation. The remaining changes are 
administrative in nature and thus none of the proposed changes create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed LCO provides a reasonable AOT to troubleshoot, repair, 
and test an inoperable ESF logic subsystem. The remaining ESF logic 
subsystem is still operable as verified by surveillance testing and 
capable of both automatic and manual ESF equipment actuation. With an 
inoperable ESF logic subsystem, the ESF logic system would not be 
single failure proof for a brief period of time. However, it is OPPD's 
position that making repairs while the plant is at power and stable is 
preferable to imposing a transient (manual shutdown) on the plant at a 
time when the ESF logic system is no longer single failure proof. 
Therefore, OPPD concludes that the proposed LCO and supporting 
administrative changes do not result in a significant reduction in a 
margin of safety.
    Based on the above considerations, it is OPPD's position that this 
proposed amendment does not involve significant hazards considerations 
as defined by 10 CFR 50.92 and the proposed changes will not result in 
a condition which significantly alters the impact of the Station on the 
environment. Thus, the proposed changes meet the eligibility criteria 
for categorical exclusion set forth in 10 CFR 51.22(c)(9) and pursuant 
to 10 CFR 51.22(b) no environmental assessment need be prepared.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: William H. Bateman.

Philadelphia Electric Company, Docket No. 50-352, Limerick 
Generating Station, Unit 1, Montgomery County, Pennsylvania

    Date of amendment request: January 12, 1998.
    Description of amendment request: The Philadelphia Electric Company 
submitted a Technical Specifications (TS) Change Request, requesting an 
amendment to the TS (Appendix A) of Operating License No. NPF-39 for 
Limerick Generating Station (LGS), Unit 1. This proposed change will 
revise TS Table 4.4.6.1.3-1 to change the withdrawal schedule for the 
first capsule to be withdrawn from 10 Effective Full Power Years (EFPY) 
to 15 EFPY.
    A revision to TS Surveillance Requirement 4.4.6.1.4 is also 
proposed. This revision will remove the references to flux wire removal 
and analysis that was originally required following the first cycle of 
operation. The referenced flux wires were never located following the 
first cycle of operation. This TS Surveillance Requirement will be 
changed to refer to the flux wires that are located within the 
surveillance capsules, which will be removed and analyzed in accordance 
with the surveillance capsule removal schedule located in TS Table 
4.4.6.1.3-1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes do not increase the probability of occurrence 
of an accident previously evaluated in the safety analysis report and 
do not affect any accident initiators as described in the SAR [Safety 
Analysis Report]. The changes revise the withdrawal schedule for the 
reactor vessel material surveillance capsules from 10 Effective Full 
Power Years (EFPY) to 15 EFPY. The capsules are not an initiator of any 
previously analyzed accident nor does the withdrawal schedule of the 
surveillance capsule affect the probability or consequences of any 
previously analyzed accident.
    These changes will not affect the Pressure-Temperature (P-T) limits 
as given in LGS Technical Specification (TS) Figure 3.4.6.1-1 and UFSAR 
[Updated Final Safety Analysis Report] Figure 5.3-4. P-T limits are 
imposed on the reactor coolant system to ensure that adequate safety 
margins exist during normal operation, anticipated operational 
occurrences, and system hydrostatic tests. The P-T limits are related 
to the RTNDT [reference temperature], as described in ASME 
Section III, Appendix G. Changes in the fracture toughness properties 
of reactor pressure vessel (RPV) beltline materials, resulting from 
neutron irradiation and the thermal environment, are monitored by a 
surveillance program in compliance with the requirements of 10 CFR 50 
Appendix H. The effect of neutron fluence on the shift in the 
RTNDT is predicted by methods given in Regulatory Guide 
1.99, Rev. 2.
    As detailed in Attachment 3 [of the licensee's application dated 
January 12, 1998], for LGS Unit 1, the combination of low expected 
RTNDT shift for the plate material due to low predicted 
fluence and excellent material chemistry, Supplemental Surveillance 
Program (SSP) data on similar material, and the inherent margin in the 
P-T curve calculations--with the withdrawal schedule of the first 
surveillance capsule modified from 10 EFPY to 15 EFPY--will result in a 
more credible set of surveillance data while ensuring the continued 
safe operation of LGS Unit 1.
    LGS's current P-T limits were established based on adjusted 
reference temperatures developed in accordance with the procedures 
prescribed in Regulatory Guide 1.99, Rev. 2, Regulatory Position 1, 
``Surveillance Data Not Available.'' Calculation of adjusted reference 
temperature by these procedures includes a conservative base fluence 
estimate, power rerate adjustment of a 110% fluence multiplier from 
startup--instead of a 105% fluence

[[Page 6989]]

multiplier since 1R06 [Unit 1 refueing outage 6], and a margin term to 
ensure conservative, upper-bound values are used for the calculation of 
the P-T limits. Revision of the first capsule withdrawal schedule will 
not affect the P-T limits because the capsule constitutes one set of 
credible surveillance data. The curves will continue to be established 
in accordance with Regulatory Position 1 procedures.
    As per Regulatory Guide 1.99, Radiation Embrittlement of Reactor 
Vessel Materials, Revision 2, Regulatory Position 2, ``Surveillance 
Data Available,'' the collection of two or more sets of credible 
surveillance data is necessary to empirically calculate the adjusted 
reference temperature (ART). Each surveillance capsule constitutes one 
set of credible surveillance data. This calculated ART can be used to 
revise the Pressure-Temperature (P-T) curves (Technical Specification 
Figure 3.4.6.1-1). Without two or more sets of credible data, the ART 
must be calculated and the P-T curves revised, based upon the 
calculational methodologies as provided in the Regulatory Guide 1.99, 
Rev. 2, Regulatory Position 1, ``Surveillance Data Not Available.'' 
These methodologies use plant specific chemistry and fluence values to 
determine a calculated shift in RTNDT. A ``margin'' term is 
then added to obtain conservative, upper-bound values of adjusted 
reference temperature.
    The existing LGS Unit 1 P-T curves are currently valid up to 12 
EFPY. With first capsule removal at either 10 or 15 EFPY, the existing 
P-T curves will require a revision prior to reaching 12 EFPY based upon 
the calculational methodologies as contained in the Regulatory Guide 
1.99, Rev. 2, Regulatory Position 1, ``Surveillance Data Not 
Available.'' Therefore, the revision to the first capsule withdrawal 
schedule results in no impact to the calculational methodologies that 
will be used for the P-T curve revision that will be necessary to 
extend the curves beyond 12 EFPY.
    The fluence data as determined from the surveillance capsule flux 
wires at 15 EFPY will provide an accurate indication of neutron 
fluence. In accordance with Regulatory Guide 1.99, Rev. 2, Regulatory 
Position 1 methodology, data from these flux wires will permit an 
adjustment of TS Figure 3.4.6.1-1 in accordance with TS surveillance 
requirement 4.4.6.1.3, if required, and will meet the requirements of 
10 CFR 50 Appendix H and ASTM E-185.
    These changes will not affect any plant safety limits or limiting 
conditions of operation. The proposed changes will not affect reactor 
pressure vessel performance as they do not involve any physical 
changes, and LGS P-T limits will remain conservative in accordance with 
Reg. Guide 1.99, Rev. 2 requirements. The proposed changes will not 
cause the RPV or interfacing systems to be operated outside of their 
design or testing limits.
    The proposed changes do not increase the consequences of a 
malfunction of equipment important to safety previously evaluated in 
the SAR. The proposed changes do not involve any physical changes to 
equipment important to safety. The potential for RPV failure will be 
adequately assessed by the proposed withdrawal schedule. In addition, 
the results from the SSP will provide industry data that bounds the 
materials used in the LGS Unit 1 reactor pressure vessel until the data 
from the first LGS Unit 1 capsule is available. The proposed changes 
provide the same level of confidence in the integrity of the vessel.
    Therefore, the proposed TS changes do not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed changes do not create the possibility of a different 
type of accident than any previously evaluated in the SAR. The proposed 
changes will revise the withdrawal schedule for the first reactor 
pressure vessel (RPV) material surveillance capsule from 10 Effective 
Full Power Years (EFPY) to 15 EFPY. These proposed changes do not 
involve a physical modification of the design of plant structures, 
systems or components. The proposed changes will not impact the manner 
in which the plant is operated, as plant operating and testing 
procedures will not be affected by the changes. No new accident types 
or failure modes will be introduced as a result of the proposed 
changes.
    LGS's current Pressure-Temperature (P-T) limits were established 
based on adjusted reference temperatures developed in accordance with 
the procedures prescribed in Regulatory Guide 1.99, Rev. 2, Regulatory 
Position 1, ``Surveillance Data Not Available.'' Calculation of 
adjusted reference temperature by these procedures includes a 
conservative base fluence estimate, power rerate adjustment of a 110% 
fluence multiplier from startup--instead of a 105% fluence multiplier 
since 1R06, and a margin term to ensure conservative, upper-bound 
values are used for the calculation of the P-T limits. Revision of the 
first capsule withdrawal schedule will not affect the P-T limits 
because the capsule constitutes one set of credible surveillance data. 
The curves will continue to be established in accordance with 
Regulatory Position 1 procedures.
    The existing LGS Unit 1 P-T curves are currently valid up to 12 
EFPY. With first capsule removal at either 10 or 15 EFPY, the existing 
P-T curves will require a revision, prior to reaching 12 EFPY, based 
upon the calculational methodologies as contained in the Regulatory 
Guide 1.99, Rev. 2, Regulatory Position 1, ``Surveillance Data Not 
Available.''
    Therefore, the Technical Specification (TS) revision to the first 
capsule withdrawal schedule results in no impact to the calculational 
methodologies that will be used for the P-T curve revision that will be 
necessary to extend the curves beyond 12 EFPY.
    The fluence data as determined from the surveillance capsule flux 
wires at 15 EFPY will provide an accurate indication of neutron 
fluence. In accordance with Regulatory Guide 1.99, Rev. 2, Regulatory 
Position 1 methodology, data from these flux wires will permit an 
adjustment of TS Figure 3.4.6.1-1 in accordance with TS Surveillance 
Requirement 4.4.6.1.3, if required, and will meet the requirements of 
10 CFR 50 Appendix H and ASTM E-185.
    The potential for reactor pressure vessel (RPV) failure will 
continue to be adequately assessed by the proposed withdrawal schedule. 
As detailed in Attachment 3, the combination of the low expected shift 
for the plate material, SSP data on similar material, and the inherent 
margin in the P-T curve calculations will result in a credible set of 
surveillance data, while ensuring the continued safe operation of LGS 
Unit 1. The proposed changes provide the same level of confidence in 
the integrity of the RPV.
    Therefore, the proposed TS changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed TS changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes to the Technical Specifications (TS) do not 
reduce the margin of safety as defined in the Bases for any TS. The 
proposed changes will not affect any safety limits, limiting safety 
system settings, or limiting conditions of operation. The proposed 
changes do not represent a change in initial conditions, system 
response time, or in any other parameter

[[Page 6990]]

affecting the accident analyses supporting the Bases of any TS. The 
proposed changes do not involve revision of the P-T limits but rather a 
revision of the withdrawal schedule for the first surveillance capsule. 
The current P-T limits were established based on the adjusted reference 
temperatures for vessel beltline materials calculated in accordance 
with Regulatory Position 1 of Reg. Guide 1.99, Rev. 2. P-T limits will 
continue to be revised as necessary for changes in adjusted reference 
temperature due to changes in fluence according to Regulatory Position 
1 until two or more credible surveillance data sets become available. 
When two or more credible surveillance data sets become available, P-T 
limits will be revised as prescribed by Regulatory Position 2 of Reg. 
Guide 1.99, Rev. 2 or other NRC approved guidance.
    The current P-T limit curves are inherently conservative and 
provide sufficient margin to ensure the integrity of the reactor 
pressure vessel. The proposed changes do not adversely affect these 
curves. The fluence data as determined from the surveillance capsule 
flux wires at 15 EFPY will provide an accurate indication of neutron 
fluence.
    In accordance with Regulatory Guide 1.99, Rev. 2, Regulatory 
Position 1 methodology, data from these flux wires will permit an 
adjustment of TS Figure 3.4.6.1-1 in accordance with TS Surveillance 
Requirement 4.4.6.1.3, if required, and will meet the requirements of 
10 CFR 50 Appendix H and ASTM E-185.
    Therefore, the proposed TS changes do not involve a reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.
    Attorney for licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, PA 19101.
    NRC Project Director: John F. Stolz.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, 
Limerick Generating Station, Units 1 and 2, Montgomery County, 
Pennsylvania

    Date of amendment request: September 2, 1997.
    Description of amendment request: This proposed Technical 
Specification (TS) Change Request revises TS Sections 4.0.5, and Bases 
Sections B 4.0.5 and B 3/4.4.8, for Limerick Generating Station (LGS), 
Units 1 and 2, pertaining to the surveillance requirement associated 
with Inservice Inspection (ISI) and Inservice Testing (IST) activities 
for American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel (B&PV) Code, Class 1, 2, and 3 components.
    The existing wording in TS Section 4.0.5, and Bases Sections B 
4.0.5 and B 3/4.4.8, stipulates that ISI and IST surveillance 
activities for ASME Code Class 1, 2, and 3 components be conducted in 
accordance with the requirements of Section XI of the ASME Code as 
required by 10 CFR 50.55a(g). The proposed changes will revise the 
applicable TS sections to only make reference to 10 CFR 50.55a, since 
the current regulations have separated the specific requirements for 
ISI and IST into sections 50.55a(g) and 50.55a(f), respectively.
    The existing wording of TS Section 4.0.5, and Bases Sections B 
4.0.5 and B 3/4.4.8, also requires that ISI and IST surveillance 
activities be conducted in accordance with the requirements of Section 
XI of the ASME Boiler and Pressure Vessel Code, except where specific 
written relief has been granted by the NRC. This wording precludes the 
immediate implementation of alternative testing in the event that a 
Code required inspection has been identified as clearly impractical. 
The proposed TS changes will revise the applicable TS sections to 
eliminate the requirement that written relief be obtained prior to 
implementation of alternative testing during the initial 120-month 
inspection interval, and the initial 12 months of subsequent intervals 
in cases where the Code required inspections have been found to be 
clearly impractical. NUREG-1482, ``Guidelines for Inservice Testing at 
Nuclear Power Plants,'' discusses impracticality as being a situation 
where a test cannot be performed due to limitations in design (which 
includes prohibitive dose rates), construction, or system 
configuration.
    Furthermore, TS Section 4.0.5b. currently discusses the required 
frequency of ISI and IST surveillance activities required by the ASME 
Code. The existing TS address testing frequencies of up to one (1) 
year. In some cases, the ASME Code requires that testing be performed 
on a two (2) year frequency. The proposed TS changes will also revise 
the TS to include a reference for tests that are conducted on a 
biennial frequency. Inclusion of this reference will permit the 
application of TS 4.0.2 criteria for ISI and IST surveillance 
activities. This will permit a 25 percent time extension to be applied 
to the surveillance frequency, if necessary, in order to allow for 
consideration of plant operating conditions when scheduling ISI and IST 
surveillance tests.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed TS changes are administrative in nature and do not 
make physical modifications or changes to the plant structures, 
systems, or components (SSC). Plant SSC will continue to function as 
designed. The proposed TS changes will not alter equipment operational 
practices or procedures.
    In the event that an ASME Section XI Code required inspection or 
test is found to be impractical due to unforeseen conditions, written 
relief would still be requested from the NRC in accordance with 
established procedures. No code required inspection will be eliminated 
from the ISI or IST Programs until written approval has been granted by 
the NRC as required [by] 10CFR50.55a. It is anticipated that the only 
time this provision would be utilized would be in the event that an 
inspection or test is discovered to be impossible or impractical to 
perform due to unforeseen or unexpected high radiation conditions, or 
physical limitations. This change will also clarify the applicability 
of surveillance intervals to biennial tests or examinations.
    The proposed TS changes will remove the inconsistencies between the 
LGS TS and the requirements of 10CFR50.55a, and will also ensure that 
the implementation of the LGS ISI and IST Programs are consistent with 
current NRC guidance as specified in NUREG-1482 and NUREG-1433, 
Revision 1.
    Therefore, the proposed TS changes do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. The proposed TS changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated.

[[Page 6991]]

    The proposed changes apply to the administrative requirements for 
testing of plant systems. No physical modifications to systems or 
components are involved. No new failure modes which could cause or 
contribute to the cause of an accident are being introduced.
    The proposed TS changes will remove the inconsistencies between the 
LGS TS and the requirements of 10CFR50.55a, and will also ensure that 
the implementation of the LGS ISI and IST Programs are consistent with 
current NRC guidance as specified in NUREG-1482 and NUREG-1433, 
Revision 1.
    Therefore, the proposed TS changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed TS changes do not involve a significant reduction 
in a margin of safety.
    No physical plant modifications or operational procedure changes 
are being made as a result of the proposed TS changes. The proposed TS 
changes apply to the ISI and IST Programs' surveillance requirements 
and do not modify the scope or frequency of these Programs as required 
by 10 CFR 50.55a. The proposed TS changes will eliminate 
inconsistencies between current TS wording and the requirements 
specified in 10CFR50.55a. In addition, the proposed changes are 
consistent with the guidance stipulated in NUREG-1482 and NUREG-1433, 
Revision 1. No physical plant modifications or operational procedure 
changes are being introduced as a result of this proposed TS Change.
    Therefore, the proposed TS changes do not involve a reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, PA 19101.
    NRC Project Director: John F. Stolz.

Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: October 8, 1997.
    Description of amendment request: This amendment proposes revisions 
to the actions to be taken in the event multiple control rods are 
inoperable.
    Basis for proposed no significant hazards 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 number and distribution of inoperable control rods is not a 
precursor to any accident, therefore the probability of an accident is 
not affected. The proposed changes assure the assumptions used in 
evaluation of accidents are satisfied, therefore there will be no 
increase in the consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    Changing the allowable number and distribution of inoperable 
control rods and the power level at which these limits apply to be 
consistent with the accident analyses does not create the possibility 
of a new or different kind of accident.
    3. Involve a significant reduction in a margin of safety because:
    The proposed changes assure the assumptions used in the accident 
analyses are satisfied, therefore there will be no affect on the margin 
of safety as a result of these changes.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Director.

Southern California Edison Company, et al., Docket Nos. 50-361 and 
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, 
San Diego County, California

    Date of amendment requests: November 6, 1995, as supplemented by 
letter dated January 9, 1998. The supplemental submittal supersedes the 
staff's proposed no significant hazards consideration determination 
evaluation for the requested changes that were published on April 10, 
1996 (61 FR 15996).
    Description of amendment requests: In the November 6, 1995, letter, 
the licensee proposed to revise Technical Specification (TS) 3.5.1, 
``Safety Injection Tanks,'' to extend, in general, the allowed outage 
time (AOT) for a single inoperable safety injection tank (SIT) from 1 
hour to 24 hours. Additionally, the licensee proposed to extend the SIT 
AOT from 1 hour to 72 hours if a single SIT becomes inoperable due to 
malfunctioning SIT water level and/or nitrogen cover pressure 
instrumentation. The January 9, 1998, letter modifies the original 
request by adding a new TS 5.5.2.14, ``Configuration Risk Management 
Program,'' that ensures a proceduralized probabilistic risk assessment-
informed process is in place that assesses the overall impact of plant 
maintenance on plant risk.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The Safety Injection Tanks (SITs) are passive components in the 
Emergency Core Cooling System (ECCS). The SITs are not accident 
initiators in any accident previously evaluated. Therefore, this change 
does not involve an increase in the probability of an accident 
previously evaluated.
    The SITs are designed to mitigate the consequences of Loss of 
Coolant Accidents (LOCAs). The proposed changes do not affect any of 
the assumptions used in deterministic LOCA analysis. Therefore, the 
consequences of accidents previously evaluated do not change.
    To fully evaluate the SIT Completion Time extension, Probabilistic 
Safety Analysis (PSA) methods were utilized. The results of these 
analyses show no significant increase in core damage frequency. As a 
result, there would be no significant increase in the consequences of 
an accident previously evaluated.
    The proposed change pertaining to SIT inoperability based solely on 
instrumentation malfunction does not involve a significant increase in 
the consequences of an accident as evaluated and endorsed by the 
Nuclear Regulatory Commission (NRC) in

[[Page 6992]]

NUREG-1366, ``Improvements to Technical Specifications Surveillance 
Requirements.''
    The Configuration Risk Management Program is an Administrative 
Program that assesses risk based on plant status. Adding the 
requirement to implement this program for Technical Specification 3.5.1 
does not affect the probability or the consequences of an accident.
    Therefore, this change does not involve a significant increase in 
the probability or consequences of any accident previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    This proposed change does not change the design, configuration, or 
method of operation of the plant. Therefore, this change does not 
create the possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed changes do not affect the limiting conditions for 
operation or their bases that are used in the deterministic analyses to 
establish the margin of safety. PSA evaluations were used to evaluate 
these changes. These evaluations demonstrate that the changes are 
either risk neutral or risk beneficial.
    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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, Irvine, California 92713.
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770.
    NRC Project Director: William H. Bateman.

Southern California Edison Company, et al., Docket Nos. 50-361 and 
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, 
San Diego County, California

    Date of amendment requests: November 8, 1995, as supplemented by 
letter dated January 9, 1998. The supplemental submittal supersedes the 
staff's proposed no significant hazards consideration determination 
evaluation for the requested changes that were published on April 10, 
1996 (61 FR 15996).
    Description of amendment requests: In the November 8, 1995, letter, 
the licensee proposed to revise Technical Specification (TS) 3.5.2, 
``ECCS--Operating,'' to extend the allowed outage time from 72 hours to 
7 days for a single low pressure safety injection train. The January 9, 
1998, letter modifies the original request by adding a new TS 5.5.2.14, 
``Configuration Risk Management Program,'' that ensures a 
proceduralized probabilistic risk assessment-informed process is in 
place that assesses the overall impact of plant maintenance on plant 
risk.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The Low Pressure Safety Injection (LPSI) system is a part of the 
Emergency Core Cooling System (ECCS) subsystem. Inoperable LPSI 
components are not considered to be accident initiators. Therefore, 
this change does not involve an increase in the probability of an 
accident previously evaluated.
    The LPSI system is primarily designed to mitigate the consequences 
of a large Loss of Coolant Accident (LOCA). This proposed change does 
not affect any of the assumptions used in the deterministic LOCA 
analysis. Therefore, the consequences of accidents previously evaluated 
do not change.
    To fully evaluate the LPSI Completion Time extension, Probabilistic 
Safety Analysis (PSA) methods were utilized. The results of these 
analyses show no significant increase in core damage frequency. As a 
result, there would be no significant increase in the consequences of 
an accident previously evaluated.
    The Configuration Risk Management Program is an Administrative 
Program that assesses risk based on plant status. Adding the 
requirement to implement this program for Technical Specification 3.5.2 
does not affect the probability or the consequences of an accident.
    Therefore, this change does not involve a significant increase in 
the probability or consequences of any accident previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    This proposed change does not change the design, configuration, or 
method of operation of the plant. Therefore, this change does not 
create the possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed change does not affect the limiting conditions for 
operation or their bases that are used in the deterministic analyses to 
establish the margin of safety. PSA evaluations were used to evaluate 
these changes.
    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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, Irvine, California 92713.
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770.
    NRC Project Director: William H. Bateman.

Southern California Edison Company, et al., Docket Nos. 50-361 and 
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, 
San Diego County, California

    Date of amendment requests: July 29, 1996.
    Description of amendment requests: The licensee proposes to revise 
Technical Specification (TS) 3.7, ``Plant Systems,'' and TS 4.3, ``Fuel 
Storage,'' to permit an increase in the licensed storage capacity of 
the spent fuel pools.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    In the course of previous analyses and the analyses required to 
support the consolidation and storage of spent fuel assemblies 
generated by the San Onofre Nuclear Generating Station Units 1, 2 and 3 
(SONGS 1, 2 and 3), the

[[Page 6993]]

enveloping scenarios described below have been considered. The limiting 
event or accident is considered that which produces the greatest 
radiological dose consequences.
    (1) Design Basis Fuel Handling Accidents. Postulated fuel handling 
accidents consider drops of either a spent fuel assembly or a 
consolidated fuel canister in the spent fuel pool (SFP) or cask pool. 
In addition to damage to the dropped fuel assembly or consolidated fuel 
canister, a fuel assembly or consolidated fuel canister seated in the 
SFP or the cask pool may be impacted by the drop. Alternatively, the 
dropped assembly or canister may fall over an empty rack cell, or fall 
onto the pool floor/liner. These various scenarios have been 
considered.
    The reference fuel in the analysis presented below is SONGS 2 and 3 
fuel. Due to the longer decay time, lower burnup, and lower operating 
power of SONGS 1 fuel, the consequences of damage to SONGS 1 fuel are 
bounded by the consequences of damage to SONGS 2 and 3 fuel.
    (a) Dropped Fuel Assembly. The limiting and design basis fuel 
assembly drop event is a 254-inch drop of a vertically-oriented fuel 
assembly, which has decayed for 72 hours, onto the SFP floor, followed 
by rotation of the fuel assembly to the horizontal position. The 
postulated bounding event results in a total of 60 fuel rods failing, 
which will not change as a result of fuel consolidation.
    The probability of a spent fuel assembly drop during movement of 
spent fuel is slightly increased by fuel consolidation because the 
candidate fuel assemblies are moved from their individual rack cell 
location to the cask pool for consolidation. However, this increase in 
probability is not significant since the process and equipment used to 
move fuel assemblies will not be changed. Additionally, fuel movement 
activities will be performed by personnel trained, qualified, and 
certified in fuel handling operations. Therefore, the increase in 
probability of a spent fuel assembly drop due to fuel consolidation is 
not significant.
    The SFP water leakage consequences of a fuel assembly drop are 
bounded by the consequences of a postulated empty spent fuel rack drop. 
The resulting leakage (approximately 49 gallons per minute) is well 
within the makeup water supply capability (150 gallons per minute). 
Additionally, the water loss would be contained within the spent fuel 
pool leak chase system and would not be released to the soil or the 
environment.
    Spent fuel assemblies will be decayed (subcritical) at least 72 
hours prior to being moved and at least 6 months prior to being 
consolidated. Administrative controls will require that fuel assemblies 
being moved to and from the consolidation work station, and when in the 
work station, be separated by more than 12 inches of water from edge to 
edge to maintain neutronic isolation. Criticality calculations show 
that with 1800 parts per million (ppm) minimum boron concentration in 
the SFP water (Technical Specifications limit of 1850 ppm includes 50 
ppm measurement uncertainty), a dropped fuel assembly event will not 
result in fuel criticality.
    Without crediting filtration by the fuel handling building (FHB) 
post-accident cleanup units, the offsite doses which result from this 
scenario are well within the required limits, i.e., less than 25 
percent (%) of the limits imposed by 10 CFR 100. The control room doses 
meet 10 CFR 50, Appendix A, General Design Criterion (GDC) 19 limits 
when crediting the control room emergency air cleanup system. 
Therefore, the consequences of a fuel handling accident remain 
enveloped by the fuel assembly drop event.
    In conclusion, the probability and consequences of a fuel assembly 
drop event will not be significantly increased by the proposed fuel 
consolidation activity.
    (b) Dropped Consolidated Fuel Canister. A dropped consolidated fuel 
canister event does not involve significantly new failure mechanisms 
compared with a dropped fuel assembly event. The limiting event in this 
category is a 74-inch drop of a consolidated fuel canister from the 
spent fuel handling machine (SFHM) into a rack cell containing a 
consolidated fuel canister. The structural integrity of the racks would 
not be impacted and both consolidated fuel canisters would remain 
intact. However, it is conservatively assumed that all 944 fuel rods 
within the two canisters (472 rods/canister  x  2 canisters) are 
damaged.
    The probability of a consolidated fuel canister drop is not 
expected to vary significantly from that expected for a fuel assembly 
drop because the methods and equipment used to move consolidated fuel 
canisters will not be significantly different from those used for fuel 
assemblies. Additionally, effective training methods, administrative 
controls, and equipment design will be developed to minimize the 
likelihood of dropping a canister during the consolidation process.
    The SFP water leakage consequences of a consolidated fuel canister 
drop are bounded by the consequences of a postulated empty spent fuel 
rack drop as discussed previously in Item 1.1(a).
    The criticality calculations show that, with the required 1800 ppm 
boron concentration in the SFP and cask pool water, there are no 
criticality consequences of postulated consolidated fuel canister 
drops. In all cases, the structural integrity of the racks will be 
maintained. The portions of the canisters where fuel is contained 
(above and inclusive of the bottom plate) will maintain their 
structural integrity in all drop cases.
    The offsite doses which result from this scenario are bounded by 
the fuel assembly drop event discussed previously in Item 1.1(a) (60 
failed fuel rods in an assembly which has decayed 72 hours) and are 
well within (less than 25% of) the limits imposed by 10 CFR 100. The 
control room doses meet the GDC 19 limits when crediting the control 
room emergency air cleanup system. Therefore, the consequences of a 
consolidated fuel canister drop event are enveloped by the limiting 
fuel assembly drop event.
    In conclusion, the probability and consequences of the limiting 
fuel drop event will not be significantly increased by storing 
consolidated fuel in canisters.
    (2) Spent Fuel Pool (SFP) Gate Drop. The limiting case is a SFP 
gate drop on a fuel assembly. Analysis has shown that only one assembly 
would be impacted and all 236 rods in the assembly potentially damaged 
subsequent to a drop of the SFP gate. The radiological consequences are 
shown to be acceptable (less than 25% of 10 CFR 100 limits).
    Current gate lift height restrictions (no more than 30 inches above 
the racks) will be maintained for fuel consolidation. With these 
restrictions, fuel in only one rack cell (either a spent fuel assembly 
with 236 rods or a consolidated fuel canister with 472 rods) would be 
impacted with all rods in the fuel assembly or canister being 
potentially damaged.
    The probability of a SFP gate drop is not significantly increased 
by fuel consolidation because the process and equipment used to move 
the gate will not change and because the gate will be kept open and not 
moved or removed when fuel is located in the cask pool during 
consolidation (administrative control).
    Despite the additional fuel rods in a consolidated fuel canister 
(472 rods versus 236 rods in a fuel assembly), the minimum six month 
decay time allows more than 99.9% of the radioactive gases to decay. 
Thus, a gate drop that results in a damaged fuel assembly 72

[[Page 6994]]

hours after shutdown is more limiting than a gate drop that results in 
a damaged consolidated fuel canister. With the analysis demonstrating 
impact of fuel in only one cell, offsite doses remain well within (less 
than 25% of) the limits of 10 CFR 100 without taking credit for the FHB 
filters. The control room emergency air cleanup system will maintain 
control room doses within GDC 19 limits.
    Therefore, the probability and consequences of a gate drop will not 
be significantly increased due to the proposed fuel consolidation 
activity.
    (3) Test Equipment Skid Drop. Current test equipment skid height 
restrictions (no more than 72 inches above rack cells containing SONGS 
2 and 3 fuel assemblies or 30 feet 8 inches above those containing 
SONGS 1 assemblies) will be maintained after fuel consolidation is 
implemented. These restrictions will ensure that the potential depth of 
penetration of test equipment skid into the racks is not sufficient to 
damage stored fuel.
    The probability of a test equipment skid drop is not affected by 
fuel consolidation because the methods and equipment used to move the 
skid will not change. In addition, there are no adverse criticality 
consequences of a test equipment skid drop on a fuel assembly or 
consolidated fuel canister, since the structural configuration of the 
fuel or of the impacted storage rack cells is not significantly changed 
because of the drop impact.
    Since no fuel is damaged, the probability and consequences of a 
test equipment skid drop will not be significantly increased due to the 
proposed fuel consolidation activity.
    (4) Cask Handling Crane Load Drops. The types of loads currently 
lifted by the cask handling crane include spent fuel casks, 
transshipment casks, and the crane load block. To support consolidation 
activities, lifts of the fuel consolidation equipment will also be 
performed by the cask handling crane. The travel path of the cask 
handling crane does not extend over spent fuel in the SFP. 
Administrative controls will prohibit operation of the cask handling 
crane, including the crane load block, within ten feet of the edge of 
the cask pool when fuel is present in the cask pool during 
consolidation. The handling of heavy loads by the cask handling crane 
is governed by the SONGS heavy loads program which has received Nuclear 
Regulatory Commission (NRC) approval. The movement of fuel 
consolidation equipment by the cask handling crane will be evaluated 
under the heavy loads program. Thus, an accident resulting from cask 
handling crane load drops into the SFP or onto irradiated fuel in the 
cask pool is not credible.
    It is expected that the consolidation work station in the cask pool 
will be temporarily removed prior to any spent fuel cask, transshipment 
cask, or other load lifts/movements over the cask pool. Other than 
insertion and removal of the consolidation work station, the equipment 
and procedures used to lift and move cask handling crane loads will be 
unaffected by fuel consolidation.
    Therefore, the probability and consequences of a spent fuel cask or 
transshipment cask drop are not significantly increased by the proposed 
fuel consolidation activity.
    (5) Mispositioning of a Consolidated Fuel Canister. The probability 
of mispositioning a consolidated fuel canister is expected to be 
comparable to that for mispositioning of a spent fuel assembly because 
the methods and equipment used to move and position consolidated fuel 
canisters in rack cells will not be significantly different from those 
used for fuel assemblies. Additionally, fuel movement activities are 
and will continue to be performed by personnel trained, qualified, and 
certified in fuel handling operations.
    The potential consequences of a mispositioned consolidated fuel 
canister relate to fuel criticality. The burnup of the fuel stored in 
the SFP before, during, and after consolidation will conform to the 
criteria provided in the Technical Specifications. With the minimum 
required 1800 ppm (1850 ppm plus 50 ppm measurement uncertainty) boron 
concentration in the SFP and the Region II racks loaded with fuel which 
meets the burnup criteria of Technical Specification 3.7.18, k-eff 
remains less than 0.90 for a consolidated fuel canister mispositioned 
in the Region II racks.
    Therefore, the probability and consequences of mispositioning a 
consolidated fuel canister are not significantly higher than the 
probability and consequences of mispositioning a fuel assembly.
    (6) Maximum Flow Blockage to Cool Spent Fuel. Flow blockage to a 
consolidated fuel canister may be caused by either damage to the 
canister or loose material in the spent fuel pool or cask pool. 
Canisters will be inspected prior to being placed in the cask pool 
(prior to loading with fuel), and if damaged during movement or 
placement in the spent fuel pool. Additionally, the existing foreign 
material exclusion control in the spent fuel pool area will be utilized 
for fuel consolidation. Therefore, the probability of blocking flow to 
a consolidated fuel canister will not be significantly increased.
    The temperature effects of a postulated flow blockage of a 
consolidated fuel canister were evaluated relative to the anticipated 
maximum cladding temperature of 700 degrees Fahrenheit (700 deg.F) 
during reactor full power. Each rack storage cell has large or multiple 
flow holes to virtually eliminate the possibility that all flow in a 
cell would be blocked by debris or foreign material. The flow openings 
in the canisters will be designed to maintain a clear flow area of at 
least 20% under all postulated blockage conditions. For the postulated 
80% flow blockage, the resulting maximum cladding temperature is 
233.1 deg.F, which is well below the maximum anticipated cladding 
temperature of 700 deg.F during reactor full power.
    Therefore, the probability and consequences of flow blockage will 
not be significantly increased by the proposed fuel consolidation 
activity.
    (7) Loss of Spent Fuel Pool (SFP) Cooling. The probability of loss 
of SFP cooling is not affected by fuel consolidation because the 
existing SFP cooling system will perform its design function without 
modification.
    The overall design basis (maximum abnormal) heat load will be 
increased due to an increased number of spent fuel elements stored. The 
cask pool may be used for temporary storage of spent fuel assemblies 
during consolidation. Loss of cooling flow to the cask pool has not 
been specifically analyzed. However, because of administrative controls 
which limit the amount of fuel permitted in the cask pool during 
consolidation and require the gate between the cask pool and the SFP to 
be open when fuel is present in the cask pool, this accident scenario 
is bounded by the SFP boiling case discussed below.
    An analysis of loss of SFP cooling has been performed using the 
design basis consolidated fuel heat load. This analysis shows that, 
without crediting the FHB filters, the offsite doses will remain well 
within (less than 25% of) the 10 CFR 100 limits. Since the reactivity 
will decrease with increasing temperature at 0 ppm boron concentration, 
there will be no adverse criticality effects. Additionally, the normal 
makeup sources to the SFP will continue to maintain adequate inventory 
and flow capacity (150 gallons per minute or gpm) to compensate for 
evaporative losses due to boiling (<112 gpm="" maximum).="" the="" temperature="" effects="" of="" sfp="" boiling="" on="" the="" sfp="" liner="" plate="" and="" concrete="" [[page="" 6995]]="" structure="" have="" been="" determined="" to="" be="" acceptable.="" therefore,="" the="" probability="" and="" consequences="" of="" a="" loss="" of="" sfp="" cooling="" event="" will="" not="" be="" significantly="" increased="" by="" the="" proposed="" fuel="" consolidation="" activity.="" (8)="" consolidation="" work="" station="" accidents.="" fuel="" consolidation="" will="" require="" additional="" fuel="" handling="" operations.="" however,="" since="" the="" fuel="" handling="" methods="" and="" equipment="" will="" not="" be="" significantly="" different="" from="" those="" currently="" used,="" consolidation="" work="" station="" accidents="" will="" be="" similar="" to="" fuel="" handling="" accidents="" already="" discussed="" in="" this="" safety="" analysis="" (dropped="" fuel="" assembly,="" dropped="" consolidated="" fuel="" canister,="" or="" other="" load="" drops).="" to="" avoid="" a="" significant="" increase="" in="" the="" probability="" of="" any="" of="" these="" accidents,="" personnel="" training="" methods,="" equipment="" design,="" and="" administrative="" controls="" will="" be="" utilized.="" administrative="" controls="" will="" require="" a="" minimum="" decay="" time="" of="" six="" months="" for="" spent="" fuel="" prior="" to="" its="" movement="" into="" the="" cask="" pool="" for="" consolidation.="" this="" restriction="" ensures="" that="" the="" limiting="" radiological="" offsite="" and="" control="" room="" dose="" consequences="" from="" a="" work="" station="" accident="" remain="" bounded="" by="" a="" fuel="" assembly="" drop.="" the="" results="" are="" well="" within="" (less="" than="" 25%="" of)="" 10="" cfr="" 100="" and="" meet="" gdc="" 19="" dose="" limits.="" fuel="" assemblies="" in="" the="" work="" station="" shall="" be="" separated="" by="" more="" than="" 12="" inches="" of="" water="" from="" edge="" to="" edge="" to="" maintain="" neutronic="" isolation="" (administrative="" control).="" the="" total="" spent="" fuel="" which="" will="" be="" permitted="" in="" the="" cask="" pool="" at="" any="" given="" time="" is="" 553="" fuel="" rods="" (administrative="" control).="" this="" quantity="" of="" fuel="" is="" equivalent="" to="" two="" full="" songs="" 2="" or="" 3="" fuel="" assemblies="" plus="" a="" damaged="" fuel="" rod="" storage="" canister="" or="" basket="" containing="" up="" to="" 81="" fuel="" rods.="" a="" criticality="" analysis="" has="" shown="" that,="" in="" the="" worst="" case="" scenario,="" at="" 1800="" ppm="" (technical="" specification="" limit="" of="" 1850="" ppm="" includes="" 50="" ppm="" measurement="" uncertainty)="" boron="" concentration,="" k-eff="" will="" be="" below="" 0.95.="" additional="" administrative="" controls="" will="" be="" imposed="" to="" ensure="" that="" a="" minimum="" of="" 400="" fuel="" rods="" or="" non-fuel="" rods="" will="" be="" loaded="" into="" a="" songs="" 2="" or="" songs="" 3="" consolidated="" fuel="" canister="" and="" a="" minimum="" of="" 324="" fuel="" rods="" or="" non-fuel="" rods="" will="" be="" loaded="" into="" a="" songs="" 1="" consolidated="" fuel="" canister.="" the="" canisters="" shall="" be="" designed="" for="" storage="" of="" fuel="" rods="" within="" a="" maximum="" allowed="" rod="" pitch.="" for="" canisters="" not="" fully="" loaded,="" the="" rod="" pitch="" shall="" be="" maintained="" by="" restraints="" inserted="" within="" the="" canister="" to="" ensure="" against="" rod="" displacement="" during="" canister="" movement="" (administrative="" control).="" these="" limitations="" ensure="" that="" the="" k-eff="" for="" a="" loaded="" consolidated="" fuel="" canister="" will="" not="" exceed="" 0.95="" with="" zero="" ppm="" boron="" concentration,="" considering="" worst="" case="" pitch="" between="" consolidated="" rods.="" with="" 1800="" ppm="" boron="" concentration="" in="" the="" pool,="" k-eff="" will="" be="" below="" 0.88="" for="" the="" worst="" case="" canister="" pitch="" between="" rods.="" thus,="" there="" are="" no="" adverse="" criticality="" consequences="" since="" the="" minimum="" number="" of="" rods="" consolidated="" in="" a="" canister="" is="" administratively="" controlled="" and="" sfp="" and="" cask="" pool="" boron="" concentration="" will="" be="" maintained="" at="" or="" above="" 1800="" ppm="" during="" consolidation.="" therefore,="" the="" consequences="" of="" a="" consolidation="" work="" station="" accident="" are="" not="" significantly="" increased="" as="" a="" result="" of="" the="" proposed="" fuel="" consolidation="" activity.="" (9)="" seismic="" events.="" the="" probability="" of="" occurrence="" of="" a="" seismic="" event="" is="" unaffected="" by="" the="" proposed="" fuel="" consolidation="" activity.="" the="" consequences="" of="" a="" design="" basis="" earthquake="" (dbe)="" have="" been="" analyzed,="" and="" the="" fuel="" consolidation="" process="" and="" consolidated="" fuel="" canisters="" will="" not="" affect="" the="" ability="" of="" the="" racks="" to="" maintain="" their="" required="" design="" basis="" function="" during="" and="" after="" a="" dbe.="" the="" spent="" fuel="" racks="" are="" designed,="" and="" the="" consolidated="" fuel="" canisters="" will="" be="" designed,="" to="" seismic="" category="" i="" requirements,="" and="" the="" consolidation="" equipment="" will="" be="" designed="" to="" seismic="" category="" ii/i="" requirements="" as="" defined="" by="" nrc="" regulatory="" guide="" 1.29,="" revision="" 3.="" the="" consolidation="" process="" provides="" the="" capability="" to="" store="" more="" spent="" fuel="" (up="" to="" approximately="" 2867="" fuel="" assemblies)="" than="" previously="" approved="" by="" the="" nrc="" (up="" to="" 1542="" fuel="" assemblies)="" in="" the="" sfp.="" the="" fuel="" handling="" building="" and="" the="" sfp="" and="" cask="" pool="" structures="" have="" been="" evaluated="" for="" the="" increased="" loading="" from="" fully-loaded="" consolidated="" fuel="" canisters="" and="" the="" loads="" found="" to="" be="" within="" the="" design="" allowables.="" thus,="" the="" probability="" or="" consequences="" of="" a="" seismic="" event="" are="" not="" significantly="" increased="" by="" the="" proposed="" fuel="" consolidation="" activity.="" (10)="" consolidated="" fuel="" canister="" stuck="" in="" a="" spent="" fuel="" rack.="" the="" probability="" of="" a="" consolidated="" fuel="" canister="" being="" stuck="" in="" a="" spent="" fuel="" rack="" is="" not="" known="" from="" experience="" since="" fuel="" consolidation="" demonstration="" projects="" conducted="" to="" date="" have="" not="" reported="" this="" type="" of="" occurrence.="" however,="" the="" canisters="" will="" be="" designed="" to="" be="" handled="" by="" the="" spent="" fuel="" handling="" machine="" (sfhm),="" will="" have="" the="" same="" approximate="" cross-sectional="" dimensions="" as="" spent="" fuel="" assemblies,="" and="" similar="" handling="" equipment="" and="" methods="" will="" be="" used.="" therefore,="" the="" failure="" mechanisms="" are="" expected="" to="" be="" comparable="" to="" those="" for="" a="" stuck="" fuel="" assembly.="" on="" this="" basis,="" the="" probability="" of="" a="" consolidated="" fuel="" canister="" being="" stuck="" in="" a="" spent="" fuel="" rack="" is="" estimated="" to="" be="" comparable="" to="" that="" for="" a="" stuck="" fuel="" assembly.="" the="" canisters="" will="" be="" designed="" to="" accommodate="" all="" operational="" and="" handling="" loads.="" a="" design="" requirement="" will="" be="" imposed="" that="" the="" canisters="" be="" capable="" of="" withstanding="" the="" maximum="" sfhm="" lift="" load="" of="" 6000="" pounds="" and="" remain="" intact="" with="" no="" fuel="" spillage.="" this="" is="" consistent="" with="" the="" criteria="" utilized="" previously="" during="" sfp="" reracking="" for="" the="" spent="" fuel="" racks="" and="" a="" jammed="" fuel="" assembly.="" with="" these="" design="" criteria="" and="" restrictions,="" deformation="" of="" rack="" cell="" geometry="" would="" not="" be="" sufficient="" to="" exceed="" the="" criticality="" acceptance="" criterion="">0.95). 
Therefore, the consequences of a stuck consolidated fuel canister would 
be bounded by the consequences of a stuck fuel assembly.
    Therefore, there is no significant increase in the probability or 
consequences of an accident previously evaluated due to the proposed 
fuel consolidation activity.
    (11) Limiting Component Cooling Water (CCW) System Heat Load 
Effects on Spent Fuel Pool Cooling. The maximum calculated heat load 
for the CCW system occurs during a Loss of Coolant Accident (LOCA). The 
probability of a LOCA, and therefore the probability of maximum heat 
load being imposed on the CCW system, is not affected by fuel 
consolidation. The reason is that spent fuel handling operations in the 
SFP or the cask pool are not, of themselves, LOCA initiators. For the 
purposes of assessing the heat load on the CCW system, the LOCA is 
divided into two phases, ``safety injection'' and ``recirculation.''
    During the safety injection phase, the SFP heat load is isolated 
from the CCW system. During the recirculation phase, CCW system cooling 
to the SFP may be reestablished manually. The recirculation phase 
represents the highest design heat load for the CCW system. Considering 
the limiting consolidated fuel heat load contribution from the SFP 
(assuming a minimum of 60 days decay of the most recent half-core 
discharged into the SFP), the CCW system has adequate capacity to still 
remove its design basis heat load.
    Therefore, the probability or consequences of a limiting design 
basis heat load event on the CCW system are not significantly increased 
by the proposed fuel consolidation activity.
    Therefore, operation of the facility in accordance with this 
proposed change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new or

[[Page 6996]]

different kind of accident from any accident previously evaluated.
    The proposed change will allow the consolidation of San Onofre 
Units 1, 2 and 3 spent fuel in canisters and the storage of these 
canisters along with fuel assemblies in the Units 2 and 3 spent fuel 
pools. Fuel consolidation is similar in nature to fuel reconstitution 
within a fuel assembly since individual rods are manipulated in both 
processes. Accidents involving consolidated fuel canisters are similar 
in nature to fuel assembly handling accidents since both use similar 
fuel handling processes and equipment. Administrative controls will be 
instituted to provide assurance that postulated events involving 
consolidated fuel will be enveloped by the spectrum of design basis 
fuel handling accidents. Furthermore, heavy load drops during spent 
fuel handling operations are accidents that have been previously 
evaluated. Additional evaluations have been performed to demonstrate 
that when the minimum boron concentration requirements of the Technical 
Specifications have been met, the criticality criterion is satisfied 
for all postulated accidents.
    Therefore, operation of the facility in accordance with the 
proposed change will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The issue of ``margin of safety,'' when applied to spent fuel 
consolidation and storage, includes the following areas:
(1) Nuclear criticality,
(2) Thermal-hydraulics,
(3) Mechanical, material and structural aspects, and
(4) Offsite doses.
    These four areas are addressed below.
    (1) Nuclear Criticality. The margin of safety that has been 
established for nuclear criticality is that, including all 
uncertainties, there is a 95% probability at a 95% confidence level 
that the effective neutron multiplication factor (k-eff) in spent fuel 
pools shall be less than or equal to 0.95, under all normal and 
postulated accident conditions. This margin of safety has been adhered 
to in the criticality analyses for fuel consolidation and the storage 
of consolidated fuel canisters.
    Criticality of fuel assemblies and consolidated fuel canisters in 
fuel storage racks is prevented by the rack design which precludes 
interactions between two fuel assemblies or two consolidated fuel 
canisters or between a fuel assembly and a consolidated fuel canister. 
This is accomplished by fixing the minimum separation between storage 
cells containing fuel assemblies or consolidated fuel canisters, using 
Boraflex, a neutron absorbing material, and utilizing strict 
administrative controls.
    During the consolidation process, fuel rods which cannot be 
consolidated will be placed in a damaged fuel rod canister or basket. 
Fuel assemblies, consolidated fuel canisters, and damaged fuel rod 
canisters or baskets moving to and from the consolidation work station 
or present in the work station shall be separated by more than 12 
inches of water, measured edge to edge, to ensure that they are 
neutronically isolated (administrative control). The total spent fuel 
which will be permitted in the cask pool at any give time is 553 fuel 
rods (administrative control). This quantity of fuel is equivalent to 
two full SONGS 2 or 3 spent fuel assemblies plus 81 fuel rods in a 
damaged fuel rod canister or basket. Additionally, the rod pitch inside 
partially loaded canisters shall be maintained by restraints inserted 
within the canister to ensure against rod displacement during canister 
movement (administrative control).
    The analytical methods utilized in the criticality analyses conform 
with American National Standards Institute (ANSI) Standard N18.2-1973, 
``Nuclear Safety Criteria for the Design of Stationary Pressurizer 
Water Reactor Plants,'' Section 5.7, Fuel Handling Systems; ANSI 
Standard 57.2-1983, ``Design Objectives for LWR Spent Fuel Storage 
Facilities at Nuclear Power Stations,'' Section 6.4.2; ANSI Standard 
N16.9-1975, ``Validation of Calculational Methods for Nuclear 
Criticality Safety;'' NRC Standard Review Plan (NUREG-0800), Section 
9.1.2, ``Spent Fuel Storage''; and the NRC guidance, ``OT Position for 
Review and Acceptance of Spent Fuel Storage and Handling 
Applications,'' (April 1978), as modified (January 1979).
    The criticality analyses performed for normal conditions assume 
zero boron concentration in the SFP water and worst-case fuel 
enrichments and burnups. Most credible accident conditions will not 
result in an increase in k-eff of the spent fuel racks. However, 
accidents, such as a heavy load drop, misloading a consolidated fuel 
canister or dropping a fuel assembly, can be postulated to increase 
reactivity. For these accident conditions, the double contingency 
principle of ANSI N16.1-1975 is applied. This principle states that it 
is not required to assume two unlikely, independent events to ensure 
protection against a criticality accident. Therefore, for accident 
conditions, the presence of soluble boron in the storage pool water can 
be assumed as a realistic initial condition since the absence of boron 
would be the second unlikely event.
    Worst case accident analyses have been performed that show that 
1800 ppm of soluble boron will maintain the spent fuel pool and cask 
storage pool k-eff less than 0.95, including uncertainties, at the 
required 95%/95% probability/confidence level.
    (2) Thermal-Hydraulics. The relevant thermal-hydraulics 
considerations for determining if there is significant reduction in a 
margin of safety are: (1) maximum fuel temperature, and (2) increase in 
temperature of the water in the pool, and (3) increase in heat load 
rejection to the environment.
    Similar to the criticality analysis, the SFP decay heat load 
calculation assumes worst-case fuel loading, enrichment, and burnup. 
The calculation uses the same methodology as that used for the original 
decay heat analysis. Standard Review Plan (SRP) Section 9.1.3 criteria 
for maximum normal and maximum abnormal heat load conditions were used 
in this evaluation.
    The effect of the increased heat load has been evaluated and it has 
been shown that, under the SRP maximum normal heat load, the existing 
spent fuel pool cooling system will maintain the bulk pool water 
temperature below 145 deg.F. This value considers a single active 
failure of one spent fuel pool cooling system pump, coincident with a 
loss of offsite power, and is consistent with Standard Review Plan, 
Section 9.1.3.III.1.d. The 145 deg.F temperature represents a small 
increase in the currently approved SFP temperature of 140 deg.F. 
However, this temperature limit was very conservatively calculated, 
considering only heat losses through the spent fuel pool heat 
exchangers, and conservatively neglecting losses through evaporation to 
the spent fuel pool area, as well as conduction to the fuel handling 
building structure mass. This increase in spent fuel pool temperature 
does not represent a significant reduction in the margin of safety, 
since the affected portions of the spent fuel pool cooling system and 
other important to safety equipment in the fuel handling building are 
qualified for this slightly higher temperature and will still perform 
the necessary safety functions when required.
    A thermal-hydraulic analysis has been performed which shows that 
the maximum local water temperatures along the fuel channels will 
remain below the nucleate boiling condition values, even with the 
maximum postulated flow blockage (80%) of the consolidated fuel 
canisters. The

[[Page 6997]]

maximum calculated fuel cladding temperature for the design basis 
condition is 233.1 deg.F, which is well below the anticipated maximum 
cladding temperature of 700 deg.F during full power operation of the 
reactor.
    SONGS 2 and 3 conduct refueling by offloading either half the core 
or the full core. The full core offload refueling provides the greater 
of the two heat loads. Therefore, in addition to the SRP criteria, the 
heat load during refueling operations was also evaluated. For this case 
the heat load was evaluated assuming a two year refueling cycle, the 
spent fuel pool completely filled with consolidated fuel (except for 
the last core offload), and the full core offloaded at 150 hours of 
decay. Under these conditions, a single SFP cooling pump with two heat 
exchangers will maintain the SFP temperature below 160 deg.F, assuming 
the component cooling water temperature is 88 deg.F and the ocean water 
temperature is 76 deg.F. Thus, the SFP cooling system meets the single 
active failure criterion for the maximum refueling heat load condition.
    With the postulated SRP maximum abnormal heat load, the bulk pool 
temperature will reach a maximum of 160 deg.F with two pumps and two 
heat exchangers in operation. This maximum temperature is well below 
the SRP maximum temperature limit of 212 deg.F. Also, according to the 
SRP guidance, a single active failure need not be considered for the 
maximum abnormal heat load case.
    The shutdown cooling system (SDCS), if available, can be used as an 
alternate heat dissipation path for cooling the SFP. The SDCS has been 
evaluated for the maximum normal and maximum abnormal heat loads and it 
has been determined that the system and interconnecting ties are 
adequate to maintain the SFP temperature below 145 deg.F for the 
maximum normal heat load and below 160 deg.F for the maximum abnormal 
heat load. Since the maximum abnormal heat load bounds the maximum 
refueling heat load, there is no need to evaluate the SDCS for the 
maximum refueling heat load. For the maximum refueling heat load, the 
SDCS does not meet the single failure criterion for SFP cooling; 
however, the use of the SDCS for SFP cooling during Modes 5 and 6 of 
plant operation has previously been evaluated and considered acceptable 
by the NRC.
    The heat load rejection to the environment will only increase by 
approximately 0.03%.
    Thus, there is no significant reduction in a margin of safety, as 
determined by thermal-hydraulics considerations.
    (3) Mechanical, material, and structural aspects. The main safety 
function of the spent fuel pool and the storage racks is to maintain 
the spent fuel assemblies and consolidated fuel canisters in a safe 
configuration through normal and/or abnormal loadings. Abnormal loads 
include an earthquake, impact due to a cask drop, drop of a spent fuel 
assembly or consolidated fuel canister, or drop of a heavy load 
including a spent fuel pool gate. The mechanical, material, and 
structural design of the consolidation work station and consolidated 
fuel canisters will be in accordance with the applicable portions of 
the ``NRC OT Position of Review and Acceptance of Spent Fuel Storage 
and Handling Applications'' and other applicable NRC guidance and 
industry codes. The canisters will be designed to Seismic Category I 
requirements, and the consolidation equipment will be analyzed and 
either restrained or anchored as appropriate to meet Seismic Category 
II/I requirements as defined by NRC Regulatory Guide 1.29, Revision 3. 
The consolidation work station and consolidated fuel canister materials 
will be compatible with the spent fuel rods and spent fuel assemblies, 
and the spent fuel pool water chemistry. Therefore, margins of safety 
relative to mechanical, material, and structural aspects of the 
proposed fuel consolidation activities will not be significantly 
reduced.
    (4) Offsite and Control Room Doses. The offsite and control room 
dose consequences of accidents involving consolidated fuel canisters or 
fuel consolidation activities were evaluated. To determine the 
radiological consequences, all credible accidents related to fuel 
consolidation activities were considered. The analyses assume that 
spent fuel has decayed a minimum of 6 months prior to commencing the 
consolidation process.
    The limiting accident for fuel consolidation is a 74-inch drop of a 
consolidated fuel canister from the Spent Fuel Handling Machine (SFHM) 
onto a rack cell containing a consolidated fuel canister. Although both 
consolidated fuel canisters would remain intact, it is conservatively 
assumed that all 944 fuel rods within the two canisters (472 rods/
canister  x  2 canisters) are damaged. The resultant release of 
radioactivity, after escaping from the spent fuel pool, is exhausted 
from the fuel handling building (FHB) over a two-hour period; no credit 
for FHB isolation system or FHB filters was taken.
    The results demonstrate that, with a minimum decay time of 6 months 
and no credit taken for isolation or filtration, the radiological 
consequences of the worst case consolidated fuel accident would not 
result in releases that would exceed 25% of the 10 CFR 100 limits. The 
results also demonstrate that the control room doses would meet the 10 
CFR 50, Appendix A, GDC 19 limits when crediting the control room 
emergency air cleanup system.
    Therefore, operation of the facility according to this proposed 
change will not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, Irvine, California 92713.
    Attorney for licensee: T.E. Oubre, Esquire, Southern California 
Edison Company, P.O. Box 800, Rosemead, California 91770.
    NRC Project Director: William H. Bateman.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San 
Diego County, California

    Date of amendment requests: January 24, 1997.
    Description of amendment requests: The licensee proposes to revise 
Surveillance Requirement 3.8.1.9 to Technical Specification 3.8.1, ``AC 
Sources--Operating.'' This change will revise the surveillance 
requirement to more accurately reflect safety analysis 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. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change would revise Surveillance Requirement (SR) 
3.8.1.9 to more clearly reflect test conditions and be in greater 
agreement with NUREG 1432.
    The Voltage and Frequency limits are made tighter, to accurately 
reflect plant design requirements. Discussion regarding reactive power 
loading is eliminated from the SR, consistent with the wording of NUREG 
1432, Rev. 1, and added to the Bases.

[[Page 6998]]

    Operation of the facility would remain unchanged as a result of the 
proposed changes and no assumptions or results of any accident analyses 
are affected. Therefore, the proposed change will not involve a 
significant increase in the probability or consequences of any accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change would revise Surveillance Requirement (SR) 
3.8.1.9 to more clearly reflect test conditions and be in greater 
agreement with NUREG 1432.
    Operation of the facility would remain unchanged as a result of the 
proposed change. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed change would revise Surveillance Requirement (SR) 
3.8.1.9 to more clearly reflect test conditions and be in greater 
agreement with NUREG 1432. The Voltage and Frequency limits are made 
more restrictive, to accurately reflect the assumptions made in the 
SONGS accident analysis. Consequently, no reduction in any margin to 
safety exists.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, Irvine, California 92713.
    Attorney for licensee: T.E. Oubre, Esquire, Southern California 
Edison Company, P.O. Box 800, Rosemead, California 91770.
    NRC Project Director: William H. Bateman.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant (Farley), Units 1 and 2, Houston 
County, Alabama

    Date of amendments request: December 31, 1997.
    Description of amendments request: The proposed amendments would 
revise the Technical Specifications to change the nuclear 
instrumentation system intermediate range neutron flux reactor trip 
setpoint and allowable value.
    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 in Intermediate Range reactor trip setpoint from 
25% RTP [rated thermal power] to 35% RTP, the associated allowable 
value change, and the deletion of the redundant references to the IR 
[intermediate range] high flux and PR [power range] high flux low 
setpoints do not involve a significant increase in the probability or 
consequences of an accident previously evaluated in the Farley FSAR 
[Final Safety Analysis Report]. The IR reactor trip neither causes any 
accident nor provides primary protection for any accident in the Farley 
FSAR. No new accident initiators have been identified because of this 
proposed revision. No new performance requirements for any system that 
is used to mitigate dose consequences have been imposed by this 
proposed change. No input assumption to any dose consequence 
calculation is affected by this proposed change. All previously 
reported dose consequences remain bounding. Therefore, the radiological 
consequences to the public resulting from any accident previously 
evaluated in the FSAR have not significantly increased.
    2. The proposed Technical Specifications change to the IR reactor 
trip setpoint, associated allowable value change, and the deletion of 
the redundant references to the IR high flux and PR high flux low 
setpoints do not create the possibility of a new or different kind of 
accident from any previously evaluated in the FSAR. No new accident 
scenarios, failure mechanisms or limiting single failures are 
introduced as a result of the increase in IR setpoint from 25% RTP to 
35% RTP. No new challenges to the safety-related Reactor Trip System 
have been identified. The NIS [nuclear instrument system] hardware has 
not been modified, and Farley will continue to perform periodic IR 
channel calibration and surveillance in accordance with Technical 
Specifications. All previously identified accident scenarios remain 
bounding since the IR trip setpoint provides no primary accident 
protection. Therefore, the possibility of a new or different kind of 
accident is not created.
    3. The proposed increase in the IR reactor trip setpoint from 25% 
RTP to 35% RTP, the associated allowable value change, and the deletion 
of the redundant references to the IR high flux and PR high flux low 
setpoints do not involve a significant reduction in the margin of 
safety. All previously established acceptance limits continue to be met 
for all events, since the IR trip does not provide any primary 
protective action for any accident scenario. Changing the IR setpoint 
and allowable value will not invalidate its backup function. There are 
no physical modifications required for the protection system. This 
change will not affect the operation of any other safety-related 
equipment. Farley-specific setpoint uncertainty calculations support 
the setpoint change. Since all acceptance limits continue to be met, 
there is no significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama.
    NRC Project Director: Herbert N. Berkow.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-424 and 50-425, Vogtle Electric 
Generating Plant, Units 1 and 2, Burke County, Georgia

    Date of amendment request: January 22, 1998. The application 
supersedes, in its entirety, the application dated September 13, 1996.
    Description of amendment request: The proposed application would 
change the Vogtle Electric Generating Plant (VEGP) Technical 
Specification (TS) 3.8.1, ``AC Sources--Operating,'' as follows: (1) 
The completion time for restoration of one required offsite circuit 
would be increased from 6 to 14 days from discovery of failure to meet 
the Limiting Condition for Operation (LCO); (2) a new required action 
B.2 would be added along with the existing Condition B required actions 
for one Diesel Generator (DG) inoperable, to verify the availability of 
the Standby Auxiliary

[[Page 6999]]

Transformer (SAT) within 1 hour and once per 12 hours thereafter, and 
restore the DG to operable status within 14 days from discovery of 
failure to meet the LCO; (3) a new required action B.5.1 would be added 
to verify that the combustion turbine electrical power generation 
capability of Plant Wilson is functional and sufficiently reliable to 
provide assurance of black-start generation capability within 72 hours 
of entry into Condition B or within 72 hours prior to entry into 
Condition B; (4) a new required action B.5.2 would be added for 
utilization when the combined combustion turbine generator (CTG) 
enhanced black start reliability falls below the required criteria. 
This condition allows the option to start or run at least one of the 
CTGs at Plant Wilson within 72 hours of entry to Condition B, or prior 
to entry into Condition B for preplanned maintenance; (5) a new 
condition C is being added for when one DG is inoperable and the 
required actions and completion times of B.2 are not met, i.e. the SAT 
is not verified to be available or becomes unavailable as an offsite 
source, or the required actions and completion times of B.5 associated 
with CTG operation and/or reliability are not met, then restore the DG 
to operable status within 72 hours; and (6) other changes associated 
with TS 3.8.1 conditions, required actions, or completion times are 
only the result of re-numbering due to the addition of the new 
condition and required actions of the DG extended Allowable Out-of-
Service Time (AOT) and do not reflect a change to operating 
requirements.
    In addition, a new TS 5.5.18, ``Configuration Risk Management 
Program (CRMP),'' would be added to the Administrative section of the 
TS. This section discusses the program description and use.
    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?
    No. The DGs are used to support mitigation of the consequences of 
an accident; however, they are not considered the initiator of any 
previously analyzed accident. The use of the SAT as an additional 
offsite power source coupled with the black start generation capability 
of Plant Wilson and the use of a configuration risk management program 
will more than compensate for the risk introduced by the extended DG 
Completion Times. As such, the extension of the DG Completion Times 
will not significantly increase the probability or 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?
    No. 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?
    No. This proposed TS only affects the length of the allowed outage 
time for DGs and does not change the DG testing or maintenance 
requirements. The proposed TS still requires the DGs to be maintained 
Operable to the same standard as before. The use of the SAT as an 
additional offsite power source coupled with the black start generation 
capability of Plant Wilson and the use of a configuration risk 
management program has been shown to provide more than adequate 
compensation for the potential risk of the extended DG Completion 
times. The proposed change in DG completion times in conjunction with 
the added availability of the SAT, continue to provide adequate 
assurance of the capability to provide power to the ESF [Engineered 
Safety Features] buses. 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: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia.
    NRC Project Director: Herbert N. Berkow.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama

    Date of amendment request: December 30, 1997.
    Description of amendment request: The proposed amendment would 
change Table 3.5-1 and associated notes. The changes would remove a 
potential non-conservative operating configuration for the Residual 
Heat Removal Service Water (RHRSW) System pumps that could result in a 
loss of two pumps following a single failure of diesel-generator A or B 
thereby reducing the number of pumps available to less than the number 
required by the Final Safety Analysis Report. The changes also would 
allow (for units with fuel loaded) reducing the minimum-required number 
of RHRSW pumps by one pump for each unit that has been in cold shutdown 
for more than 24 hours. The associated Basis 3.5 also would be changed 
to reflect these changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.
    A. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated (10 CFR 
50.92(c)(1)) because the proposed changes do not involve any plant 
structures, systems, or components that are initiators of any accident 
previously evaluated, and the changes do not decrease the capability of 
the RHRSW system to transfer reactor core and emergency equipment heat 
loads to the ultimate heat sink.
    B. The changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated (10 CFR 
50.92(c)(2)) because there are no changes to plant structures, systems, 
or components, and the changes do not affect the manner by which the 
facility is operated. The proposed changes are consistent with the 
Final Safety Analysis Report analysis for the design basis accident.
    C. The changes do not involve a significant reduction in a margin 
of safety (10 CFR 50.92(c)(3)) because the proposed changes do not 
affect the manner by which the facility is operated or involve 
equipment or features which affect the operational characteristics of 
the facility. The proposed amendment would increase the diversity of 
power supplies associated with the residual heat removal cooling 
function thereby improving conformance to the single failure criterion.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff

[[Page 7000]]

proposes to determine that the amendment request involves no 
significant hazards consideration.
    Local Public Document Room location: Athens Public Library, 405 E. 
South Street, Athens, Alabama 35611.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

Vermont Yankee Nuclear Power Corporation, Docket Nos. 50-271, Vermont 
Yankee Nuclear Power Station, Windham County, Vermont

    Date of amendment request: December 11, 1997.
    Description of amendment request: The proposed amendment would 
revise the safety limit minimum critical power ratio (SLMCPR) values 
for Cycle 20 operation. The specific changes are:
    (1) Page 6, Technical Specification 1.1A. replace the cycle number 
(19) to (20) and the SLMCPR for Cycle 19 (1.10) with that for Cycle 20 
(1.11).
    (2) Page 6, Technical Specification 1.1A. replace the SLMCPR for 
Cycle 19 single loop operation (1.12) with the Cycle 20 value (1.13).
    Calculations for Vermont Yankee Nuclear Power Station (VYNPC) by 
General Electric Company have determined that the current SLMCPR values 
for single and dual loop operation contained in the Technical 
Specifications (1.10 and 1.12) are not applicable to the upcoming fuel 
cycle (Cycle 20) due to core loading design and fuel type changes. The 
Cycle 20 values are 1.11 and 1.13.
    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 basis of the Safety Limit MCPR is to ensure no mechanistic fuel 
damage is calculated to occur if the limit is not violated. The new 
SLMCPR preserves the existing margin to transition boiling and the 
probability of fuel damage is not increased. The derivation of the 
revised SLMCPR for Vermont Yankee Cycle 20 for incorporation into the 
Technical Specifications, and its use to determine cycle-specific 
thermal limits, have been performed using NRC approved methods. These 
calculations do not change the method of operating the plant and have 
no effect on the probability of an accident initiating event or 
transient.
    Based on the above, VYNPC has concluded that the proposed change 
will not result in 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 changes result only from a specific analysis for the 
Vermont Yankee Cycle 20 core reload design. These changes do not 
involve any new method for operating the facility and do not involve 
any facility modifications. No new initiating events or transients 
result from these changes.
    Based on the above, VYNPC has concluded that the proposed change 
will not create the possibility of a new or different kind of accident 
from those previously evaluated.
    3. Does this change involve a significant reduction in a margin of 
safety?
    The margin of safety as defined in the Technical Specification 
bases will remain the same. The new SLMCPR is calculated using NRC 
approved methods which are in accordance with the current fuel design 
and licensing criteria. Additionally, interim implementing procedures, 
which incorporate cycle-specific parameters, have been used. The SLMCPR 
remains high enough to ensure that greater than 99.9% of all fuel rods 
in the core will avoid transition boiling if the limit is not violated, 
thereby preserving the fuel cladding integrity.
    As a result, VYNPC has concluded that the proposed change will not 
result in 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: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Project Director: Ronald Eaton.

Previously Published Notices of Consideration of Issuance of Amendments 
to Facility Operating Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: December 5, 1997.
    Brief description of amendment: Revisions to the Crystal River Unit 
3 design basis relating to starting logic of reactor building fan 
coolers.
    Date of publication of individual notice in the Federal Register: 
January 15, 1998 (63 FR 2423).
    Expiration date of individual notice: February 17, 1998
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal River, Florida 34428.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.

[[Page 7001]]

    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see: (1) The 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: October 2, 1997.
    Brief description of amendment: The proposed change would revise 
the Updated Final Safety Analysis Report to revise the credit assumed 
for iodine decontamination by the spent fuel pool water during a 
postulated fuel handling accident.
    Date of issuance: January 27, 1998.
    Effective date: January 27, 1998.
    Amendment No.: 177.
    Facility Operating License No. DPR-23: Amendment authorizes changes 
to the facilitiy's Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: November 19, 1997 (62 
FR 61838). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 27, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of application for amendments: May 21, 1996, as supplemented 
on November 18, 1997, December 3, 1997, January 8, 1998 and January 13, 
1998.
    Brief description of amendments: The amendments relocate the 
reactor coolant system pressure and temperature limits for heatup, 
cooldown, low-temperature operation and hydrostatic testing, and the 
low-temperature overpresssure protection (LTOP) system setpoint curves 
into a Pressure Temperature Limits Report (PTLR).
    Date of issuance: January 23, 1998.
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 98, 98, 89, 89.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 18, 1997 (62 
FR 66394). The January 8, 1998 and January 13, 1998, submittals 
provided additional clarifying information that did not change the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 23, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of application for amendments: January 30, 1997, as 
supplemented by letter dated December 9, 1997. Additional information 
was submitted in ComEd's letters of May 23, 1997, August 8, 1997 and 
January 7, 1998.
    Brief description of amendments: The amendments revise the 
technical specifications and associated bases related to the primary 
containment pressure and reactor coolant system volume. The changes 
resulted from the replacement of the steam generators at Byron, Unit 1 
and Braidwood, Unit 1.
    Date of issuance: January 22, 1998.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 97, 97, 88 and 88.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 23, 1997 (62 FR 
19826) and December 19, 1997 (62 FR 66699).
    The May 23, 1997, August 8, 1997, December 9, 1997 and January 7, 
1998, letters provided additional information that did not change the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated January 22, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of application for amendments: February 18, 1997, as 
supplemented by letter dated September 22, 1997.
    Brief description of amendments: The amendments change the 
Technical Specification requirements for steam generator water level to 
support steam generator replacement at Byron, Unit 1, and Braidwood, 
Unit 1.
    Date of issuance: January 15, 1998.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 96, 96, 87 and 87.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 12, 1997 (62 FR 
11491). The September 22, 1997, submittal provided additional 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 15, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

[[Page 7002]]

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of application for amendments: June 17, 1997, as supplemented 
November 26, 1997, and January 9, 1998.
    Brief description of amendments: The amendments revise the 
technical specifications to update the containment vessel structural 
integrity surveillance requirements to meet the provisions of a recent 
revision to 10 CFR 50.55a, and to relocate details of the surveillance 
requirements to a licensee-controlled program.
    Date of issuance: January 29, 1998.
    Effective date: Effective immediately and shall be implemented 
within 60 days.
    Amendment Nos.: 99, 99, 90 and 90.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 19, 1997 (62 
FR 66697). The November 26, 1997, and January 9, 1998, letters provided 
additional clarifying information that did not change the staff's 
initial proposed no significant hazards considerations determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated January 29, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of application for amendments: September 8, 1997, as 
supplemented on January 6, 1998.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 4.5.2.b.3 and the associated Bases to bring the 
Byron, Unit 1, and Braidwood, Unit 1, requirements into conformance 
with the Unit 2 requirements that were approved on August 13, 1997. The 
revision adds a requirement to the Unit 1 TS Surveillance Requirements 
for verifying that the Chemical and Volume Control (CV) System is full 
of water every 31 days; to include ultrasonically examining the piping 
at the CV206 valve for Byron, Unit 1 (CV207 valve for Braidwood, Unit 
1), if the train B CV pump is idle. The revision also removes the 
condition that the Unit 1 requirements will be applicable only until 
the end of the current cycle (Unit 1-Cycle 8 for Byron, and Unit 1-
Cycle 7 for Braidwood). The amendments affect Unit 2 only in that the 
units share common TS.
    As an administrative action by the NRC that only involves the 
format of the licenses and does not authorize any activities outside 
the scope of the applications, the NRC has amended the Byron and 
Braidwood operating licenses to include an Appendix C, ``Additional 
Conditions,'' and added a license condition associated with the 
proposed TS changes.
    Date of issuance: January 30, 1998.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 100, 100, 91 and 91.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Facility Operating Licenses and the 
Technical Specifications.
    Date of initial notice in Federal Register: November 5, 1997 (62 FR 
59914). The January 6, 1998, submittal provided additional clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 30, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: August 12, 1997.
    Brief description of amendments: The amendments revise the LaSalle 
County Station Technical Specifications by removing Surveillance 
Requirement 4.7.1.3.c which requires that every 18 months all areas 
within the lake screenhouse be inspected to ensure that sediment has 
not been deposited to a depth greater than 1 foot.
    Date of issuance: January 23, 1998.
    Effective date: Immediately, to be implemented prior to restart 
from L1F35 for Unit 1 and prior to restart from L2RO7 for Unit 2.
    Amendment Nos.: 122 and 107.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 22, 1997 (62 FR 
54870).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 23, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver 
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

    Date of application for amendments: September 11, 1997.
    Brief description of amendments: These amendments relocate the 
reactor trip system and engineered safety feature actuation system 
reponse times from Technical Specification (TS) Tables 3.3-2 and 3.3-5 
to Section 3 of the Beaver Valley Power Station, Unit Nos. 1 and 2 
Licensing Requirements Manual (LRM) in accordance with the guidance 
provided in NRC Generic Letter 93-08. Neither the response time limits 
nor the surveillance requirements for performing response time testing 
are altered by these amendments. Any future changes to the LRM will be 
controlled in accordance with the requirements of 10 CFR 50.59. These 
amendments also make several editorial changes in TSs 3.3.1.1 and 
3.3.1.2, as well as making conforming changes to the Bases for these 
TSs.
    Date of issuance: January 20, 1998.
    Effective date: Both units, as of date of issuance, to be 
implemented within 30 days.
    Amendment Nos.: 210 and 88.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications and Appendices C (Unit No. 1) and 
D (Unit No. 2) of the Licenses.
    Date of initial notice in Federal Register: October 22, 1997 (62 FR 
54871).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 20, 1998.

[[Page 7003]]

    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: June 14, 1997, supplemented 
August 4, September 2, 17, 25, November 5, 15, 19, 21, December 3, 5, 
11, 24, 1997, January 15, and 22, 1998.
    Brief description of amendment: Changes to Technical Specification 
(TS) relating to small break loss of coolant accident mitigation, 
emergency diesel generator (EDG) upgrade and EDG load rejection test 
and steady state loads.
    Date of issuance: January 24, 1998.
    Effective date: January 24, 1998.
    Amendment No.: 163.
    Facility Operating License No. DPR-72: Amendment revised the TS.
    Date of initial notice in Federal Register: October 8, 1997 (62 FR 
52581). The letters dated August 4, September 2, 17, 25, November 5, 
15, 19, 21, December 3, 5, 11, 24, 1997, and January 15, and 22, 1998, 
provided clarifying information that did not change the initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 24, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No.3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: December 1, and 13, 1997 and 
January 19, 1998.
    Brief description of amendment: Revise License Condition 2.C.(5) to 
delete the requirement relating to installation and testing of flow 
indicators in the emergency core cooling system to provide indication 
of 40 gallons per minute flow for boron dilution.
    Date of issuance: January 27, 1998.
    Effective date: January 27, 1998.
    Amendment No.: 164.
    Facility Operating License No. DPR-72: Amendment revises License 
Condition 2.C.(5) and adds a new License Condition 2.C.11.
    Date of initial notice in Federal Register: November 12, 1997 (62 
FR 60733). Letters dated December 1 and 13, 1997 and January 19, 1998 
provided supplemental information which did not affect the original no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 27, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone 
Nuclear Power Station, Unit No. 3, New London County, Connecticut

    Date of application for amendment: November 14, 1997.
    Brief description of amendment: The amendment changes Technical 
Specification 4.5.2.d.1 to clarify the wording and increase the 
setpoint for the open pressure interlock.
    Date of issuance: January 23, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 156.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 17, 1997 (62 
FR 66138).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 23, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: January 9, 1995, as supplemented by 
letters dated October 17, 1996, and January 26, 1998.
    Brief description of amendment: The amendment revises the technical 
specifications by deleting toxic gas monitoring requirements for all 
chemicals except ammonia. The monitoring requirements for ammonia will 
remain in the technical specifications.
    Date of issuance: January 26, 1998.
    Effective date: January 26, 1998.
    Amendment No.: 183.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11137).
    The October 17, 1996, and January 26, 1998, supplemental letters 
provided additional clarifying information and did not change the 
staff's original no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated January 26, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.

Philadelphia Electric Company, Docket No. 50-352, Limerick Generating 
Station, Unit 1, Montgomery County, Pennsylvania

    Date of application for amendment: October 24, 1997.
    Brief description of amendment: This amendment changes Sections 
3.1.3.6 and 4.1.3.6 of the Unit 1 Technical Specifications to allow 
operation of control rod 50-27, uncoupled from its driver, for the 
remainder of Cycle 7. The amendment specifies conditions under which 
control rod 50-27 may be operated and modifies existing surveillance 
requirements to verify control rod position by use of neutron 
instrumentation.
    Date of issuance: January 16, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 124.
    Facility Operating License No. NPF-39: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 19, 1997 (62 
FR 61844).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 16, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.

Power Authority of the State of New York, Docket No. 50-286, Indian 
Point Nuclear Generating Unit No. 3, Westchester County, New York

    Date of application for amendment: September 8, 1997, as 
supplemented November 3, 1997.

[[Page 7004]]

    Brief description of amendment: The requested amendment modifies 
the f(I) function. The f(delta I) function is defined in the 
TS as a function of the indicated difference between the top and bottom 
detectors of the power range nuclear ion chambers. This function is 
used in the calculation of the overtemperature delta T (OTdelta T) 
reactor trip.
    Date of issuance: January 26, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 177.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 22, 1997 (62 FR 
54876). The November 3, 1997, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 26, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa.

Public Service Electric & Gas Company, Docket No. 50-272, Salem Nuclear 
Generating Station, Unit No. 1, Salem County, New Jersey

    Date of application for amendment: December 11, 1997.
    Brief description of amendment: The amendment provides a one-time 
change to the Technical Specifications to allow purging of the 
containment during Modes 3 (Hot Standy) and 4 (Hot Shutdown) upon the 
return to power from the current refueling outage (1R13).
    Date of issuance: January 29, 1998.
    Effective date: As of the date of issuance, to be implemented 
within seven days.
    Amendment No.: 206.
    Facility Operating License No. DPR-70: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 18, 1997 (62 
FR 66397).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 29, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of application for amendments: October 21, 1997.
    Brief description of amendments: These amendments revise the 
Technical Specifications to extend the Modes from 1 and 2 that the 
Reactor Trip System Power Range Nuclear Instrumentation--low setpoint 
is to be operable to Modes 1, 2, and 3, when the reactor trip breakers 
are in the closed position and the control drive system is capable of 
rod withdrawl.
    Date of issuance: January 29, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment Nos.: 205 and 187.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 19, 1997 (62 
FR 68146).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 29, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama

    Date of amendments request: June 30, 1997, as supplemented 
September 25, 1997.
    Brief Description of amendments: The amendments change the 
Technical Specifications to incorporate requirements necessary to 
change the basis for prevention of criticality in the fuel storage 
pool. The change eliminates the credit for Boraflex as a neutron 
absorbing material in the fuel storage pool criticality analysis.
    Date of issuance: January 23, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1-133; Unit 2-125.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: August 27, 1997 (62 FR 
45464).
    The staff found that the supplement did not change the conclusions 
of the proposed no significant hazards consideration; therefore, 
renotification of the Commission's proposed determination of no 
significant hazards consideration was not necessary.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 23, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.

    Dated at Rockville, Maryland, this 4th day of February 1998.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 98-3269 Filed 2-10-98; 8:45 am]
BILLING CODE 7590-01-P