[Federal Register Volume 65, Number 105 (Wednesday, May 31, 2000)]
[Notices]
[Pages 34743-34755]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-13518]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from May 6, 2000, through May 19, 2000. The last 
biweekly notice was published on May 17, 2000 (65 FR 31354).

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 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By June 30, 2000, 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 electronically from 
the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room). 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

[[Page 34744]]

petition must satisfy the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. 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 electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

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: April 18, 2000.
    Description of amendment request: The amendments would revise 
Technical Specifications (TS) 3.7.10 and 3.7.12 for Catawba Units 1 and 
2. The proposed changes address degraded pressure boundaries on the 
Auxiliary Building Filtered Ventilation Exhaust System and the Control 
Room Area Ventilation System. The proposed changes in TS 3.7.10 and 
3.7.12 would add Notes which allow the affected ventilation system 
boundaries to be opened intermittently under administrative controls. 
Also, it would add a new condition in TS 3.7.10 and 3.7.12. This new 
condition will require that the boundaries for these two systems be 
returned to an operable status within 24 hours, when both trains of 
these systems are inoperable due to an inoperable boundary.
    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:

    In accordance with the criteria set forth in 10 CFR 50.91 and 
50.92, Duke Energy Corporation has evaluated this license amendment 
request and determined it does not represent a significant hazards 
consideration. The following is provided in support of this 
conclusion.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The Control Room Area Ventilation System (CRAVS), Control 
Room pressure boundary, the Auxiliary Building Filtered Ventilation 
Exhaust System (ABFVES), or the Emergency Core Cooling System (ECCS) 
pump rooms area pressure boundary are not assumed to be initiators 
of any analyzed accident. Therefore, the proposed changes contained 
in this LAR [license amendment request] have no significant impact 
on the probability of occurrence of any previously analyzed 
accident.
    The proposed new condition for the CRAVS and ABFVES Technical 
Specifications (TS) would permit a 24-hour period to take action to 
restore an inoperable pressure boundary to OPERABLE status. The 
consequences of implementing the 24 hour Completion Time are 
reasonable based upon: (1) The low probability of a design basis 
accident occurring during this time period, (2) additional actions 
that are available to the operator to minimize doses (e.g., self 
contained breathing apparatus and alternate control room air 
intakes), and (3) the availability of an operable CRAVS/ABFVES train 
to provide a filtered environment (albeit with potential unfiltered 
leakage).
    For cases where any of the affected control room or ECCS pump 
room area/pump rooms pressure boundaries are opened intermittently 
under administrative controls, appropriate compensatory measures 
would be required by the proposed TS to ensure the pressure boundary 
can be rapidly restored. Based on the compensatory measures 
available to the plant operators and the administrative controls 
required to rapidly restore an opened pressure boundary, the 
accident consequences do not cause an increase in dose above the 
applicable General Design Criteria, Standard Review Plan, or 10 CFR 
100 limits. The plant operators will continue to maintain the 
ability to mitigate a design basis event.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. No changes are being made to actual plant hardware which 
will result in any new accident causal mechanisms. Also, no changes 
are being made to the way in which the plant is being operated. 
Therefore, no new accident causal mechanisms will be generated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    No. Margin of safety is related to the ability of the fission 
product barriers to perform their design functions during and 
following accident conditions. These barriers include the fuel 
cladding, the reactor coolant system, and the containment system. 
The performance of these barriers will not be significantly degraded 
by the proposed changes. The proposed changes would allow affected 
pressure boundaries to be degraded for a limited period of time (24 
hours).

[[Page 34745]]

However, the probability of a design basis event occurring during 
this time is low and additional actions (e.g., breathing apparatus) 
would also be taken to minimize dose to the plant operators. When 
the boundaries are open on an intermittent basis, as permitted by 
the changes proposed in this LAR, administrative controls would be 
in place to ensure that the integrity of the pressure boundaries 
could be rapidly restored. Therefore, it is expected that the plant, 
and the operators, would maintain the ability to mitigate design 
basis events and none of the fission product barriers would be 
affected by this change. Therefore, the proposed change is not 
considered to result in a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: Richard L. Emch, Jr.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio

    Date of amendment request: April 5, 2000.
    Description of amendment request: The proposed amendment implements 
technical specification (TS) changes associated with thermal-hydraulic 
stability monitoring. New TS 3.3.1.3, ``Oscillation Power Range Monitor 
(OPRM) Instrumentation,'' will provide the minimum operability 
requirements for the OPRM channels, the Required Actions when they 
become inoperable, and appropriate surveillance requirements. The OPRMs 
will provide automatic ``detect and suppress'' actions to replace the 
administrative controls currently in effect through operator training 
and manual actions. The amendment would remove monitoring guidance from 
TS 3.4.1, ``Recirculation Loops Operating,'' that will no longer be 
necessary due to the activation of the automatic OPRM instrumentation. 
Finally, the amendment would update TS 5.6.5, ``Core Operating Limits 
Report (COLR),'' to require the applicable setpoints for the OPRMs to 
be included in the COLR.
    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 specifies limiting conditions for operation, 
required actions and surveillance requirements for the Oscillation 
Power Range Monitor (OPRM) system, and allows operation in regions 
of the power to flow map currently restricted by the requirements of 
Interim Corrective Actions (ICAs) and certain limiting conditions of 
operation of Technical Specification (TS) 3.4.1. The restrictions of 
the ICAs and TS 3.4.1 were imposed to ensure adequate capability to 
detect and suppress conditions consistent with the onset of thermal-
hydraulic (T-H) oscillations that may develop into a T-H instability 
event. A T-H instability event has the potential to challenge the 
Minimum Critical Power (MCPR) safety limit. The OPRM system can 
automatically detect and suppress conditions necessary for T-H 
instability. With the activation of the OPRM System, the 
restrictions of the ICAs and TS 3.4.1 will no longer be required.
    The probability of a T-H instability event is impacted by power 
to flow conditions during operation inside specific regions of the 
power to flow map, in combination with power shape and inlet 
enthalpy conditions, such that only under such conditions can the 
occurrence of an instability event be postulated to occur. Operation 
in these regions may increase the probability that operation with 
conditions necessary for a T-H instability can occur. However, when 
the OPRM is OPERABLE with operating limits as specified in the Core 
Operating Limits Report (COLR), the OPRM can automatically detect 
the onset of significant local power oscillations and generate a 
trip signal. Actuation of a Reactor Protection System (RPS) trip 
will suppress conditions necessary for T-H instability and decrease 
the probability of a T-H instability event. In the event the trip 
capability of one or more of the OPRM channels is not maintained, 
the proposed change includes Required Actions which limit the period 
of time before the affected OPRM channel (or RPS system) must be 
placed in the tripped condition. If these actions would result in a 
trip function such as a scram, or if the OPRM trip capability is not 
maintained, an alternate method to detect and suppress thermal 
hydraulic oscillations is required, i.e., the same ICAs as are in 
place today. In either case the duration of the period of time 
allowed by the Required Actions is limited, and the probability of a 
T-H instability event during this limited time is not significantly 
increased.
    Several changes to TS 3.4.1 are made which are more consistent 
with, or conservative with, respect to the reviewed and approved 
Standard Technical Specifications for Boiling Water Reactors. These 
generic changes are considered applicable to the Perry Nuclear Power 
Plant. They simply provide guidance on the operator actions to be 
taken and the associated time limits when the Specification is 
entered, and do not impact the probability of occurrence of an 
accident. For the above reasons, the proposed change does not result 
in a significant increase in the probability of an accident 
previously evaluated.
    An unmitigated T-H instability event is postulated to cause a 
violation of the MCPR safety limit. The proposed change ensures 
mitigation of T-H instability events prior to challenging the MCPR 
safety limit if initiated from anticipated conditions, by detection 
of the onset of oscillations and actuation of an RPS trip signal. 
The OPRM also provides the capability of an RPS trip being generated 
for T-H instability events initiated from unanticipated but 
postulated conditions. These mitigating capabilities of the OPRM 
system will become available as a result of the proposed change and 
have the potential to reduce the consequences of anticipated and 
postulated T-H instability events. The OPRM installation has been 
evaluated to not adversely impact other installed equipment such as 
the Average Power Range Monitors (APRMs) or the RPS in a manner that 
could prevent response to various postulated events, so those events 
will not have increased consequences due to the OPRMs. Therefore, 
the proposed change does not significantly increase the consequences 
of an accident previously evaluated.
    Therefore, the proposed change, which specifies limiting 
conditions for operation, required actions and surveillance 
requirements for the OPRM system, and allows operation in certain 
regions of the power to flow map, does not significantly increase 
either the probability or consequences of an accident previously 
evaluated.
    2. The proposed change would not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change specifies limiting conditions for operation, 
required actions and surveillance requirements of the OPRM system, 
and allows operation in regions of the power to flow map currently 
restricted by the requirements of ICAs and TS 3.4.1. The OPRM system 
uses input signals shared with APRM and rod block functions to 
monitor core conditions and generate an RPS trip when required. 
Quality requirements for software design, testing, implementation 
and module self-testing of the OPRM system provide assurance that 
new equipment malfunctions due to software errors are not created. 
The design of the OPRM system also ensures that neither operation 
nor malfunction of the OPRM system will adversely impact the 
operation of other systems and no accident or equipment malfunction 
of these other systems could cause the OPRM system to malfunction or 
cause a different kind of accident. Therefore, operation with the 
OPRM system does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    Operation in regions currently restricted by the requirements of 
ICAs and TS 3.4.1 is within the nominal operating domain and ranges 
of plant systems and components, and within the range for which 
postulated accidents have been evaluated. Therefore operation within 
these regions does not

[[Page 34746]]

create the possibility of a new or different kind of accident from 
any accident previously evaluated. The changes to TS 3.4.1 to be 
more consistent, or conservative, with respect to the reviewed and 
approved Standard Technical Specifications, simply provide guidance 
on the operator actions to be taken and the associated time limits 
when the Specification is entered, and also do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Therefore, the proposed change, which specifies limiting 
conditions for operation, required actions and surveillance 
requirements of the OPRM system, and allows operation in certain 
regions of the power to flow map, does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change will not involve a significant reduction 
in the margin of safety.
    The proposed change specifies limiting conditions for operation, 
required actions and surveillance requirements of the OPRM system 
and allows operation in regions of the power to flow map currently 
restricted by the requirements of ICAs and TS 3.4.1.
    The OPRM system monitors small groups of LPRM [local power range 
monitor] signals for indication of local variations of core power 
consistent with T-H oscillations, and generates an RPS trip when 
conditions consistent with the onset of oscillations are detected. 
An unmitigated T-H instability event has the potential to result in 
a challenge to the MCPR safety limit. The OPRM system provides the 
capability to automatically detect and suppress conditions which 
might result in a T-H instability event, and thereby maintains the 
margin of safety by providing automatic protection for the MCPR 
safety limit while reducing the burden on the control room 
operators. Therefore, operation with the OPRM system does not 
involve a significant reduction in a margin of safety. In the event 
an OPRM channel becomes inoperable, the proposed change includes 
actions which limit the period of time before the affected OPRM 
channel (or RPS system) must be placed in the tripp[ed] condition. 
If these actions would result in a trip function such as a scram (or 
if the OPRM trip capability is not maintained), the alternate method 
to detect and suppress thermal hydraulic oscillations (the current 
ICAs) is required to be put in place. The duration of the period of 
time allowed by the Required Actions is limited, and the probability 
of a significant T-H instability event during this limited time is 
not significantly increased.
    Operation in regions currently restricted by the requirements of 
ICAs and Technical Specification [TS] 3.4.1 is within the nominal 
operating domain and ranges of plant systems and components, and 
within the range assumed for initial conditions considered in the 
analysis of anticipated operational occurrences and postulated 
accidents. Therefore, operation in these regions does not involve a 
significant reduction in the margin of safety. The changes to TS 
3.4.1 to be more consistent, or conservative, with respect to the 
reviewed and approved Standard Technical Specifications, simply 
provide guidance on the operator actions to be taken and the 
associated time limits when the Specification is entered, and also 
do not significantly reduce the margin of safety.
    Therefore, the proposed change, which specifies limiting 
conditions for operation, required actions and surveillance 
requirements of the OPRM system, and allows operation in certain 
regions of the power to flow map, 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.
    Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

Florida Power and Light Company, et al., Docket No. 50-335, St. Lucie 
Plant, Unit No. 1, St. Lucie County, Florida

    Date of amendment request: April 23, 2000.
    Description of amendment request: The proposed license amendment 
(PLA) is associated with the required timing for containment hydrogen 
recombiner post operation insulation resistance testing. This PLA 
revises Unit 1 Technical Specification 3/4.6.4.2, Electric Hydrogen 
Recombiners--W, to clarify the requirement for the post-operation 
insulation resistance test of Surveillance Requirement 4.6.4.2.b.4.
    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 does not involve an increase in the 
probability or consequences of any accident previously evaluated. 
This PLA provides a clarification of the Technical Specification 
surveillance requirements for verifying hydrogen recombiner 
operability and reliability. This PLA has no affect on the testing 
requirements, test frequency, or acceptance criteria for recombiner 
operability. This change allows vendor recommended guidance and in-
house methodology to be established when conducting recombiner 
heater resistance testing. This will enable consistency in testing 
and will allow trending for determination of the material condition 
of the recombiner heaters. The PLA change provides clarification and 
preserves the intent of the basis to monitor the material condition 
of the recombiner heaters. Additionally, this change provides 
consistency and is identical with the Unit 2 Technical Specification 
surveillance. As such, this change is considered administrative in 
nature.
    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 will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. This PLA is considered administrative in nature and will 
not alter the way in which the hydrogen recombiner is operated or 
tested. This PLA allows vendor recommended guidance to be 
established in order to perform consistent testing and to allow 
meaningful trending of the results to verify recombiner operability. 
This PLA has no affect on the testing requirements, test frequency, 
or acceptance criteria for recombiner operability. This PLA does not 
result in any plant configuration changes or new failure modes.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed amendment does not involve a reduction in the 
margin of safety. This administrative PLA clarifies the surveillance 
requirement of the subject Technical Specification by allowing the 
establishment of vendor recommendations and in-house testing 
methodology to provide consistent testing conditions and allow 
meaningful trending of results. This PLA has no affect on the 
testing requirements, test frequency, or acceptance criteria for 
recombiner operability. As such, the assumptions and conclusions of 
the accident analyses in the UFSAR [Updated Final Safety Analysis 
Report] remain valid and the associated safety limits will continue 
to be met.

    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.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

Florida Power and Light Company, Docket No. 50-251 and 50-252, Turkey 
Point Units 3 and 4 in Miami-Dade County

    Date of amendment request: April 27, 2000.
    Description of amendment request: Florida Power and Light Company 
(FPL)

[[Page 34747]]

requests to amend the Turkey Point Unit 3 Facility Operating License 
DPR-31 Fire Protection license condition 3.G, and to amend the Turkey 
Point Unit 4 Facility Operating License DPR-41 Fire Protection license 
condition 3.F. The proposed revisions to the Facility Operating 
Licenses are required to incorporate references to NRC Safety 
Evaluations issued in support of 10 CFR 50 Appendix R granted 
exemptions. In addition, the proposed amendments requests to modify 
Appendix A of the Facility Operating Licenses DPR-31 and DPR-41 of the 
Turkey Point Units 3 and 4 Technical Specifications (TS), Section 
4.7.6.g. Due to an oversight, the submittal for the request of License 
Amendments Nos. 201 and 195 for Section 6.0 ``Administrative 
Controls,'' L-99-056, dated March 8, 1999, discussed revision to TS 
Section 4.7.6.g on TS Page 3/4 7-21, but inadvertently did not attach 
the revised marked up Page 3/4 7-21.
    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 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments do not involve an increase in the 
probability or consequences of an accident previously evaluated 
because the proposed changes are administrative in nature adding 
references to exemptions granted by the NRC and to reflect 
relocation of record retention requirements from the TS to the 
UFSAR. These amendments will not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because they do not affect assumptions contained in plant safety 
analyses, the physical design and/or operation of the plant, nor do 
they affect Technical Specifications that preserve safety analysis 
assumptions. Therefore, the proposed changes do not affect the 
probability or consequences of accidents previously analyzed.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed changes to the Facility Operating Licenses and the 
Technical Specifications are administrative in nature and can not 
create the possibility of a new or different kind of accident from 
any previously evaluated since the proposed amendments will not 
change the physical plant or the modes of plant operation defined in 
the facility operating license. No new failure mode is introduced 
due to the administrative changes since the proposed changes do not 
involve the addition or modification of equipment nor do they alter 
the design or operation of affected plant systems, structures, or 
components.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The operating limits and functional capabilities of the affected 
systems, structures, and components are unchanged by the proposed 
amendments. The proposed changes to the Facility Operating License 
Conditions and the TS are administrative in nature and do not reduce 
any of the margins 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.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook 
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: April 28, 2000.
    Description of amendment request: The Seabrook Station Technical 
Specifications (TSs) are proposed to be revised to implement the 
Relaxed Axial Offset Control (RAOC) operating strategy in support of 
the use of upgraded Westinghouse fuel with Intermediate Flow Mixers.
    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:
    The proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes to TS 2.1.1, 3.2.1, 4.2.1.1, 4.2.2.2, 4.2.2.3, 
4.2.2.4, 6.8.1.6.b, and changes to the aforementioned TS Bases, are in 
support of North Atlantic's long-term operating strategy to refuel and 
operate, commencing with Cycle 8, with Biweekly Notice Coordinator 
upgraded Westinghouse fuel with Intermediate Flow Mixers (VANTAGE+(w/
IFMs)). Evaluations/analyses of accidents which are potentially 
affected by the parameters and assumptions associated with the fuel 
upgrade and RAOC strategy have shown that all design standards and 
applicable safety criteria will continue to be met. The consideration 
of these changes does not result in a situation where the design, 
material, and construction standards that were applicable prior to the 
change are altered. Therefore, the proposed changes occurring with the 
fuel upgrade will not result in any additional challenges to plant 
equipment that could increase the probability of any previously 
evaluated accident.
    The proposed changes associated with the fuel upgrade and RAOC 
strategy do not affect plant systems such that their function in the 
control of radiological consequences is adversely affected. The actual 
plant configuration, performance of systems, and initiating event 
mechanisms are not being changed as a result of the proposed changes. 
The design standards and applicable safety criteria limits will 
continue to be met and therefore fission barrier integrity is not 
challenged. The proposed changes associated with fuel upgrade and RAOC 
strategy have been shown not to adversely affect the response of the 
plant to postulated accident scenarios. The proposed changes will 
therefore not affect the mitigation of the radiological consequences of 
any accident described in the Updated Final Safety Analysis Report 
(UFSAR).
    The proposed changes to TS Table 2.2-1, TS 3.2.2, TS 3.2.3, and the 
title on page 3/4 2-6 are editorial changes to correct either 
typographical errors, simplification of statements, clarification of 
specific parameters associated with temperature pressure measurements, 
making some notations consistent with improved Standard Technical 
Specifications -- Westinghouse Plants, NUREG-1431, Rev. 1, and 
relocating additional cycle-specific values for temperature, pressure 
and time constants to the [Core Operating Limits Report] COLR, or 
correcting an erroneous title. These changes do not result in a change 
to the design basis of any plant structure, system or component or 
parameters currently specified in the COLR, therefore, operation of the 
facility within the prescribed limits of TS remains unchanged.
    The proposed change to TS 3.2.1, ACTION a.2, to delete the need to 
reduce the power range neutron flux high trip setpoints subsequent to 
reducing rated thermal power (RTP) to less than 50% whenever axial flux 
difference (AFD) is outside of the applicable limits specified in the 
COLR, does not significantly increase the

[[Page 34748]]

probability or consequences of an accident previously evaluated.
    Therefore, for the reasons stated above, the probability or 
consequences of an accident previously evaluated are not significantly 
increased for all the proposed TS changes presented herein.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The possibility for a new or different type of accident from any 
accident previously evaluated is not created since the proposed changes 
associated with the fuel upgrade and RAOC strategy do not result in a 
change to the design basis of any plant structure, system or component. 
These proposed changes do not cause the initiation of any accident nor 
create any new failure mechanisms. Equipment important to safety will 
continue to operate as designed. Component integrity is not challenged. 
The proposed changes do not result in any event previously deemed 
incredible being made credible.
    The proposed changes are not expected to result in conditions that 
are more adverse and are not expected to result in any increase in the 
challenges to safety systems.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes will assure continued compliance within the 
acceptance limits previously reviewed and approved by the NRC for use 
of upgraded fuel features with RAOC. All of the appropriate acceptance 
criteria for the various analyses and evaluations will continue to be 
met.
    The proposed editorial changes do not change the current limits 
specified in Technical Specifications.
    Removing the requirement for manually reducing the power range 
neutron flux high trip setpoint does not result in a significant 
reduction in a margin of safety. There are other levels of trip 
protection to terminate a rapid rise in power excursion, such as the 
overtemperature delta-temperature (OT-T) trip function and previous 
power range neutron flux high trip setpoint. In addition, a rapid rise 
in power to greater than 50 percent RTP with AFD outside limits does 
not immediately create an unacceptable situation. The increased 
potential for a reactor trip caused by the manual manipulation of the 
setpoint needlessly exposes the plant to an unnecessary trip with the 
potential for an undesirable plant transient which may unnecessarily 
challenge safety systems.
    Therefore, the proposed aforementioned TS changes do not involve a 
signification reduction in a margin of safety.
    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.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: February 1, 2000, as supplemented by 
letter dated April 13, 2000.
    Description of amendment request: The proposed amendment proposes 
changes to the cable spreading room technical specifications to permit 
pressurizing the cable spreading room to a pressure that exceeds the 
pressure of the adjacent control room envelope area during testing.
    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:

    In accordance with 10 CFR 50.92, NNECO has reviewed the proposed 
changes and has concluded that they do not involve a significant 
hazards consideration (SHC). The basis for this conclusion is that 
the three criteria of 10 CFR 50.92(c) are not compromised. The 
proposed changes do not involve an SHC because the changes would 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed Technical Specification and Bases changes to 
exclude the requirements of Surveillance Requirements (SRs) 
4.7.7.e.2, 4.7.8.c.2, and 4.7.8.c.3 during pressurization testing of 
the Cable Spreading Room (CSR) will not increase the probability of 
an accident previously evaluated. Operation of the Control Room 
Emergency Air Filtration System and the Control Room Envelope 
Pressurization System cannot cause an accident to occur.
    The proposed Technical Specification and Bases changes to 
exclude the requirements of SRs 4.7.7.e.2, 4.7.8.c.2, and 4.7.8.c.3 
during pressurization testing of the CSR may adversely impact the 
consequences of previously evaluated accidents. During CSR 
pressurization testing, the Control Room Emergency Air Filtration 
and the Control Room Envelope Pressurization Systems may not be able 
to pressurize and maintain the Control Room envelope at a positive 
pressure with respect to adjacent areas and the outside atmosphere. 
As a result, radioactivity released from a design basis accident may 
enter the Control Room envelope. However, since the CSR area will 
actually be at a higher pressure than the outside atmosphere (during 
CSR pressurization testing), radioactive leakage into the CSR area, 
and subsequently into the Control Room envelope, should not occur 
after the temporary fan has been stopped. Administrative controls 
will be established to immediately stop the temporary fan and 
rapidly depressurize the CSR area in the event Control Building 
isolation is necessary. Once the CSR area is depressurized, the 
Control Room Emergency Air Filtration System and the Control Room 
Envelope Pressurization System will be able to function as designed 
to mitigate the consequences of the accident. In addition, the 
probability of a design basis accident (DBA) occurring while the CSR 
is pressurized is low. Therefore, exempting the requirements of SRs 
4.7.7.e.2, 4.7.8.c.2, and 4.7.8.c.3 during CSR pressurization 
testing will not result in a significant increase in the 
consequences of an accident previously evaluated.
    The proposed Technical Specification and Bases change to clarify 
the mode of operation of the Control Room Emergency Air Filtration 
System when the pressurization requirement of SR 4.7.7.e.2 applies, 
will have no adverse effect on plant operation, or the availability 
or operation of any accident mitigation equipment. The plant 
response to the design basis accidents will not change. In addition, 
the proposed change can not cause an accident. Therefore, there will 
be 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.
    The proposed Technical Specification and Bases changes to 
exclude the requirements of SRs 4.7.7.e.2, 4.7.8.c.2, and 4.7.8.c.3 
during pressurization testing of the CSR, and to clarify the mode of 
operation of the Control Room Emergency Air Filtration System when 
the pressurization requirement of SR 4.7.7.e.2 applies, will not 
alter the plant configuration (no new or different type of permanent 
equipment will be installed) or require any new or unusual operator 
actions. Temporary equipment will be utilized to pressurize the CSR, 
and administrative controls, using additional personnel beyond the 
normal shift complement, will be implemented to restore the CSR to a 
configuration that will allow the Control Room Emergency Air 
Filtration System and the Control Room Envelope Pressurization 
System to function as designed to mitigate the consequences of an 
accident. The temporary equipment and administrative controls that 
will be implemented to perform the CSR pressurization testing will 
not introduce any new failure modes that could result in a new 
accident. Also, the response of the plant and the operators 
following these accidents is unaffected by the changes. Therefore, 
the proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

[[Page 34749]]

    3. Involve a significant reduction in a margin of safety.
    The proposed Technical Specification and Bases changes to 
exclude the requirements of SRs 4.7.7.e.2, 4.7.8.c.2, and 4.7.8.c.3 
during pressurization testing of the CSR may adversely impact the 
ability of the Control Room Emergency Air Filtration System and the 
Control Room Envelope Pressurization System to function as designed 
to protect the Control Room Operators following a DBA, and to use 
other accident mitigation equipment contained within the Control 
Room envelope. However, the administrative controls that will be 
established to immediately stop the temporary fan and rapidly 
depressurize the CSR area if Control Building isolation is necessary 
will provide reasonable assurance that the habitability of the 
Control Room envelope will be maintained. Therefore, exempting the 
requirements of SRs 4.7.7.e.2, 4.7.8.c.2, and 4.7.8.c.3 during CSR 
pressurization testing will not result in a significant reduction in 
a margin of safety.
    The proposed Technical Specification and Bases change to clarify 
the mode of operation of the Control Room Emergency Air Filtration 
System when the pressurization requirement of SR 4.7.7.e.2 applies 
will have no adverse effect on plant operation, or the availability 
or operation of any accident mitigation equipment. The plant 
response to the design basis accidents will not change. Therefore, 
there will be no significant reduction in a margin of safety.
    The proposed changes do not alter the design, function, or 
operation of the equipment involved. The impact of the proposed 
changes has been analyzed, and it has been determined they do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated, do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated, and do not involve a significant reduction in a margin of 
safety. Therefore, NNECO has concluded the proposed changes do not 
involve an SHC.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: James W. Clifford.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: May 4, 2000.
    Description of amendment request: The proposed amendment would add 
new sections to the Technical Specifications (TSs) addressing missed 
surveillance test requirements and establishing a TS Bases control 
program, revise TS Chapter 6 to allow use of generic titles in lieu of 
plant-specific titles, allow an alternative when the radiation 
protection manager does not meet the qualifications of Regulatory Guide 
1.8, relocate sections of TS Chapter 6 pertaining to onsite and offsite 
review and special inspections to the Operational Quality Assurance 
Plan, and correct typographical errors.
    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.
    Operation of the Monticello plant in accordance with the 
proposed changes does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. 
None of the proposed changes involves a physical modification to the 
plant, a new mode of operation or a change to the Updated Safety 
Analysis Report transient analysis. These proposed amendments 
conform to the guidance of NUREG-1433, Revision 1, which was 
previously issued by the NRC.
    The proposed changes do not reduce the level of qualification or 
training and the aggregate knowledge of the plant staff remains 
intact. In total, these changes do 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 
analyzed.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the proposed changes do not introduce a new mode of plant 
operation, surveillance test requirement or involve a physical 
modification to the plant. These proposed amendments generally 
conform to the guidance of NUREG-1433, Revision 1, which was 
previously issued by the NRC.
    The proposed changes are administrative in nature. The changes 
propose to relocate specifications from the Technical Specifications 
to the Operational Quality Assurance Plan through which adequate 
control is maintained.
    The proposed changes do not alter the design, function or 
operation of any plant component and therefore no new accident 
scenarios are created. Therefore, the possibility of a new or 
different kind of accident from any accident previously evaluated 
would not be created by these amendments.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The proposed changes do not involve a significant reduction in a 
margin of safety because the current Technical Specification 
requirements for safe operation of the Monticello plant are 
maintained. The proposed changes are administrative in nature and do 
not involve a physical modification to the plant, a new mode of 
operation or a change to the Updated Safety Analysis Report 
transient analyses. The proposed changes do not alter the scope of 
equipment currently required to be operable or subject to 
surveillance testing nor does the proposed change affect any 
instrument setpoints or equipment safety functions.
    Therefore, a significant reduction in the margin of safety would 
not be involved with these proposed changes.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment requests: May 3, 2000 (PCN-516).
    Description of amendment requests: The amendment application 
proposes to revise the San Onofre Nuclear Generating Station, Units 2 
and 3, Technical Specification (TS) 3.4.3, ``RCS Pressure and 
Temperature (P/T) Limits,'' and the associated Bases. The proposed 
change would reduce the minimum boltup temperature from 86  deg.F to 65 
 deg.F for the reactor coolant system during the period of time when 
the reactor vessel head bolts are in tension.
    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:

    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?
    Response: No.
    This proposed change is a request to revise Technical 
Specification 3.4.3, ``Pressure Temperature Limits.'' The proposed 
change reduces the Minimum Boltup Temperature (MBT) from 86 deg.F to 
65 deg.F. During operations below 86 deg.F, the plant is in a 
shutdown mode, open to the atmosphere, and depressurized.

[[Page 34750]]

This proposed change does not affect the shape of the Pressure 
Temperature Limits when Reactor Coolant System (RCS) temperature is 
above 86 deg.F. Therefore, the probability or consequences of an 
accident previously evaluated will not be increased by operating the 
facility in accordance with this proposed change.
    Will operation of the facility in accordance with this proposed 
change create the possibility of a new or different kind of accident 
from any accident previously evaluated?
    Response: No.
    This proposed change does not change the design or configuration 
of the plant. Therefore, this proposed change will not create the 
possibility of a new or different kind of accident from any accident 
that has been previously evaluated.
    (3) Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    This proposed change involves reducing the MBT from 86 deg.F to 
65 deg.F. This proposed change meets the American Society of 
Mechanical Engineers (ASME) Code requirements for establishing the 
minimum temperature in the reactor pressure vessel flange region 
when the pressure does not exceed 20% of the pre-operational 
hydrostatic test pressure. All margins of safety established by the 
ASME Code requirements are maintained. The operation of the facility 
in accordance with 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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, 
Unit 1(WBN), Rhea County, Tennessee

    Date of amendment request: April 10, 2000 (TS 99-013).
    Description of amendment request: The proposed amendment requests 
NRC's approval to use the F* alternate repair criterion in the 
tubesheet region of the steam generator (SG). The F* criterion 
addresses the action required when degradation has been detected in the 
top of the mechanically expanded portion of SG tubes within the SG 
tubesheet.
    The proposed change designates a portion of the tube for which tube 
degradation of a defined type does not necessitate remedial action, 
except as dictated for compliance with tube leakage limits as set forth 
in the WBN Technical Specifications (TS). The proposed amendment would 
modify the TS which provide tube inspection requirements and acceptance 
criteria to determine the level of degradation for which the tube may 
remain in service. The proposed amendment would add definitions 
required for the F* alternate plugging criterion and prescribe the 
portion of the tube subject to the criterion.
    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:

    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The presence of the tubesheet enhances the tube integrity in the 
region of the hardroll by precluding tube deformation beyond its 
initial expanded outside diameter. The resistance to both tube 
rupture and tube collapse is strengthened by the presence of the 
tubesheet in that region. Hardrolling of the tube into the tubesheet 
results in an interference fit between the tube and the tubesheet. 
Tube rupture cannot occur because the contact between the tube and 
tubesheet does not permit sufficient movement of tube material. In a 
similar manner, the tubesheet does not permit sufficient movement of 
tube material to permit buckling collapse of the tube during 
postulated loss-of-coolant-accident (LOCA) loadings.
    The type of degradation for which the alternate plugging 
criterion, F*, has been developed (cracking with a circumferential 
orientation) can theoretically lead to a postulated tube rupture 
event, provided that the postulated through-wall circumferential 
crack exists near the top of the tubesheet. An evaluation including 
analysis and testing has been performed to determine the resistive 
strength of roll expanded tubes within the tubesheet. That 
evaluation provides the basis for the acceptance criteria for tube 
degradation subject to the F* criterion.
    The F* length of roll expansion is sufficient to preclude tube 
pullout from tube degradation located below the F* distance, 
regardless of the extent of the tube degradation. The existing 
technical specification leakage rate requirements and accident 
analysis assumptions remain unchanged in the unlikely event that 
significant leakage from this region does occur. As noted above, 
tube rupture and pullout are not expected for tubes using the F* 
alternate plugging criterion. Any leakage out of the tube from 
within the tubesheet at any elevation in the tubesheet is fully 
bounded by the existing steam generator tube rupture analysis 
included in the WBN Unit 1 Final Safety Analysis Report (FSAR). The 
proposed alternate plugging criterion (F*) does not adversely impact 
any other previously evaluated design basis accident.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Implementation of the proposed F* tubesheet alternate plugging 
criterion does not introduce any significant changes to the plant 
design basis. Use of the criterion does not provide a mechanism to 
result in an accident initiated outside of the region of the 
tubesheet expansion. A hypothetical accident as a result of any tube 
degradation in the expanded portion of the tube would be bounded by 
the existing tube rupture accident analysis. Tube bundle structural 
integrity and leaktightness are expected to be maintained. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The use of the tubesheet alternate plugging criterion has been 
demonstrated to maintain the integrity of the tube bundle 
commensurate with the requirements of Regulatory Guide 1.121, 
``Bases for Plugging Degraded PWR Steam Generator Tubes,'' for 
indications in the free span of tubes and the primary to secondary 
pressure boundary under normal and postulated accident conditions. 
Acceptable tube degradation for the F* criterion is any degradation 
indication in the tubesheet region, more than the F* distance below 
either the bottom of the transition between the roll expansion and 
the unexpanded tube, or the top of the tubesheet, whichever is 
lower. The safety factors used in the verification of the strength 
of the degraded tube are consistent with the safety factors in the 
American Society of Mechanical Engineering (ASME) Boiler and 
Pressure Vessel Code used in steam generator design. The F* distance 
has been verified by testing to be greater than the length of roll 
expansion required to preclude both tube pullout and significant 
leakage during normal and postulated accident conditions. Resistance 
to tube pullout is based upon the primary to secondary pressure 
differential as it acts on the surface area of the tube, which 
includes the tube wall cross-section, in addition to the inside 
diameter-based area of the tube. The leak testing acceptance 
criteria are based on the primary to secondary leakage limit in the 
technical specifications and the leakage assumptions used in the 
FSAR accident analyses.
    Implementation of the alternate tubesheet plugging criterion 
will decrease the number of tubes which must be taken out of service 
with tube plugs or repaired with sleeves. Both plugs and sleeves 
reduce the RCS flow margin; thus, implementation of the F* alternate 
plugging criterion will maintain the margin of flow that would 
otherwise be reduced in the event of increased plugging or sleeving. 
Based on the above, it is concluded that the proposed change does 
not result in a significant reduction in a loss of margin with 
respect to plant safety as defined in the FSAR or the bases of the 
WBN technical specifications.


[[Page 34751]]


    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.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

Tennessee Valley Authority (TVA), Docket No. 50-390 Watts Bar Nuclear 
Plant (WBN), Unit 1, Rhea County, Tennessee

    Date of amendment request: April 10, 2000 (TS 99-014).
    Description of amendment request: The proposed amendment would 
revise the WBN Unit 1 Technical Specification (TS) to incorporate new 
requirements associated with steam generator (SG) tube inspection and 
repair. The new requirements establish an alternate voltage based SG 
tube repair criteria at tube support plate and Flow Distribution Baffle 
plate intersections. This change is consistent with NRC Generic Letter 
95-05 ``Voltage-Based Repair Criteria for Westinghouse Steam Generator 
Tubes Affected By Outside Diameter Stress Corrosion Cracking.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Tube burst criteria are inherently satisfied during normal 
operating conditions due to the proximity of the tube support plate. 
Test data indicates that tube burst cannot occur within the tube 
support plate (TSP), even for tubes which have 100 percent through-
wall electric discharge machining (EDM) notches, 0.75 inches long, 
provided that the TSP is adjacent to the notched area. Since tube to 
tube support plate proximity precludes tube burst during normal 
operating conditions, use of the criteria must retain tube integrity 
characteristics which maintain a margin of safety of 1.43 times the 
bounding faulted condition [main steam line break (MSLB)] 
differential pressure of 2405 psig. As previously stated, the 
Regulatory Guide (RG) 1.121 criterion requiring maintenance of a 
safety factor of 1.43 times the MSLB pressure differential on tube 
burst is satisfied by \3/4\-inch diameter tubing with bobbin coil 
indications with signal amplitudes less than 
VSL = 6.03 volts, regardless of the indicated 
depth measurement. At the flow distribution baffle (FDB), a safety 
factor of 3 against the normal operating condition DP is applied; 
here a voltage of VSL = 3.81 volts satisfies 
the burst capability recommendation.
    The upper voltage repair limit (VURL) will be 
determined prior to each outage using the most recently approved NRC 
database to determine the tube structural limit (VSL). 
The structural limit is reduced by allowances for nondestructive 
examination (NDE) uncertainty (VNDE) and growth 
(VG) to establish VURL. As an example, the NDE 
uncertainty component of 20 percent and a voltage growth allowance 
of 30 percent per full power year can be utilized to establish a 
VURL of 3.71 volts for TSP indications, and 2.34 volts 
for the FDB indications. The 20 percent NDE uncertainty represents a 
square-root-sum-of-the-squares (SRSS) combination of probe wear 
uncertainty and analyst variability. The flaw growth allowance 
should be an average growth rate or 30 percent per effective full 
power year, whichever is larger. The 30 percent growth allowance 
used to determine VURL is conservative for the current 
conditions at WBN Unit 1. The most current NRC approved database, 
contained in EPRI [Electric Power Research Institute] NP-7480-L, 
Addendum 2, was used to establish the VURL values for the 
FDB and TSP intersections. Once approved by NRC, the industry 
protocol for updating the database will be followed by TVA, ensuring 
that the most current database is utilized for future applications 
of the criteria.
    Relative to the expected leakage during accident condition 
loadings, it has been previously established that a postulated MSLB 
outside of containment but upstream of the main steam isolation 
valves (MSIV) represents the most limiting radiological condition 
relative to the alternate voltage based repair criteria. In support 
of implementation of the revised repair limit, it will be determined 
whether the distribution of cracking indications at the tube support 
plate intersections during future cycles are projected to be such 
that primary to secondary leakage would result in site boundary 
doses within a fraction of the 10 CFR 100 guidelines or control room 
doses within the 10 CFR 50, Appendix A, General Design Criteria 
(GDC)-19 limit. A separate calculation has determined this allowable 
MSLB leakage limit to be 10 gallons per minute (gpm) in the faulted 
loop assuming a reactor coolant system (RCS) dose equivalent iodine 
concentration of 1.0 mCi/gm. The establishment of the 10 gpm leak 
rate value is controlled by the 0 to 2 hour offsite doses at the 
site boundary for the accident initiated iodine spike case, not 
control room dose.
    The methods for calculating the radiological dose consequences 
for this MSLB are consistent with FSAR Chapter 15 and therefore, the 
WBN licensing basis. TVA bases the calculated thyroid dose 
consequences on conversion factors from the International Commission 
on Radiation Protection (ICRP) Publication 2.
    In summary, the calculated radiological consequences of the 
exclusion area boundary and the low population zone are larger than 
previously reported for the postulated steamline break event due to 
the increased leakage and more conservative iodine spiking factors. 
However, the calculated radiological consequences remain in 
compliance with the guidelines in the Standard Review Plan, Chapter 
15, 10 CFR 50, Appendix A, GDC-19 and 10 CFR 100 reported for the 
postulated steamline break event. Therefore, it is concluded that 
the proposed changes do not result in a significant increase in the 
radiological consequences of an accident previously analyzed.
    Consistent with the guidance of Section 2.c of Generic Letter 
(GL) 95-05, the WBN Unit 1 MSLB leak rate analysis performed prior 
to returning the SGs to service may be performed based on the 
projected next end-of-cycle (EOC) voltage distribution or the actual 
measured distribution at a given outage. The method to be used for 
the first outage when outside diameter stress corrosion cracking 
(ODSCC) indication growth rates are available will be based on the 
indications found during that outage. As noted in GL 95-05, it may 
not always be practical to complete EOC calculations prior to 
returning the SGs to service. Under these circumstances, it is 
acceptable to use the actual measured bobbin voltage distribution 
instead of the projected EOC voltage distribution to determine 
whether the reporting criteria is being satisfied.
    Therefore, as implementation of the 1.0 volt voltage-based 
repair criteria at WBN Unit 1 does not adversely affect steam 
generator tube integrity and implementation is shown to result in 
acceptable radiological dose consequences, the proposed TS change 
does not result in a significant increase in the probability or 
consequences of an accident previously evaluated within the WBN 
Final Safety Analysis Report (FSAR).
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from previously analyzed.
    Implementation of the proposed steam generator tube alternate 
voltage based repair criteria (1.0 volts) does not introduce any 
significant changes to the plant design basis. Neither a single or 
multiple tube rupture event would be expected in a steam generator 
in which the repair limit has been applied (during all plant 
conditions).
    The bobbin probe voltage-based tube repair criteria of 1.0 volt 
is supplemented by: enhanced eddy current inspection guidelines to 
provide consistency in voltage normalization, a 100 percent eddy 
current inspection sample size at the tube support plate elevations, 
and rotating pancake coil (RPC) inspection requirements for the 
larger indications left in service to characterize the principal 
degradation as ODSCC.
    TVA will implement a maximum normal operating condition primary 
to secondary leakage rate limit of 600 gallons per day (gpd) total 
primary to secondary leakage and 150 gpd primary to secondary 
leakage per steam generator to help preclude the potential for 
excessive leakage during all plant conditions. The 150 gpd leakage 
limit is more restrictive than the current TS operating leakage 
limit (of 500 gpd) and is intended to provide additional margin to 
accommodate a stress corrosion crack which might grow at a greater 
than expected rate or unexpectedly extend outside the thickness of 
the tube support

[[Page 34752]]

plate. Leakage trending capability consistent with EPRI Report TR-
104788, ``Primary-to-Secondary Leak Guidelines'' has been 
implemented at WBN Unit 1.
    As steam generator tube integrity upon implementation of the 1.0 
volt repair limit continues to be maintained through in-service 
inspection and primary to secondary leakage monitoring, the 
possibility of a new or different kind of accident from any accident 
previously evaluated is not created.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The use of the voltage-based bobbin probe tube support plate 
elevation repair criteria at WBN Unit 1 maintains steam generator 
tube integrity commensurate with the criteria of Regulatory Guide 
(RG) 1.121. RG 1.121 describes a method acceptable to the NRC staff 
for meeting GDCs 14, 15, 31, and 32 by reducing the probability or 
the consequences of steam generator tube rupture. This is 
accomplished by determining the limiting conditions of degradation 
of steam generator tubing, as established by in-service inspection, 
for which tubes with unacceptable cracking should be removed from 
service. Upon implementation of the proposed criteria, even under 
the worst case conditions, the occurrence of ODSCC at the tube 
support plate elevations is not expected to lead to a steam 
generator tube rupture event during normal or faulted plant 
conditions. The EOC distribution of crack indications at the tube 
support plate elevations is confirmed to result in acceptable 
primary to secondary leakage during all plant conditions and that 
radiological consequences are not adversely impacted.
    As a preventative measure, a total of 214 tubes are excluded 
from the application of the ODSCC criteria because of the combined 
effects of loss-of-coolant-accident (LOCA) + safe shutdown 
earthquake (SSE) on the steam generator component (as required by 
GDC 2). It was determined that tube deformation or through-wall 
cracks could occur in these tubes.
    As noted previously, implementation of the tube support plate 
intersection voltage-based repair criteria will decrease the number 
of tubes which must be repaired. The installation of steam generator 
tube plugs reduces the RCS flow margin. Thus, implementation of the 
1.0 volt repair limit will maintain the margin of flow that would 
otherwise be reduced in the event of increased tube plugging.

    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.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

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 application for amendments: November 23, 1999.
    Brief description of amendments: The amendments changed Technical 
Specification 5.5.7.c.1, ``Ventilation Filter Testing.'' The testing 
criteria have been changed to be consistent with the NRC request in 
Generic Letter 99-02, ``Laboratory Testing of Nuclear-Grade Activated 
Charcoal.''
    Date of issuance: May 16, 2000.
    Effective date: May 16, 2000.
    Amendment Nos.: 209 and 237.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73086) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 16, 2000.
    No significant hazards consideration comments received: No.

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: December 22, 1999, as 
supplemented on March 1, 2000.
    Brief description of amendments: The amendments relocate Reactor 
Coolant System (RCS) related cycle-specific parameter limits from the 
technical specifications to, and thus expanding, the Core Operating 
Limits Reports (COLRs) for Byron, Units 1 and 2, and Braidwood, Units 1 
and 2.
    Date of issuance: May 15, 2000.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 113 and 106.
    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: February 23, 2000 (65 
FR 9003). The March 1, 2000, 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 May 15, 2000.
    No significant hazards consideration comments received: No.

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: July 14, 1999, as supplemented 
on January 21, February 15, February 23, March 10, March 24, March 31 
(two letters), April 7 and April 14, 2000.
    Brief description of amendments: The amendments increase the 
maximum reactor core power level to 3489 megawatts thermal; an increase 
of 5 percent of rated core thermal power, for

[[Page 34753]]

LaSalle County Station, Units 1 and 2. In addition, the proposed 
amendments correct a non-conservative value in the upper limit for 
drywell and suppression chamber internal pressure that was discovered 
during the course of the power uprate review.
    Date of issuance: May 9, 2000.
    Effective date: For Unit 1, immediately, to be implemented within 
60 days; for Unit 2, immediately, to be implemented prior to start up 
of cycle 9.
    Amendment Nos.: 140 and 125.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the license and Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46427). The letters dated January 21, February 15, February 23, March 
10, March 24, two letters on March 31, April 7, and April 14, 2000, 
contain supplemental, clarifying information that did not change the 
staff's initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 9, 2000.
    No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket No. 50-374, LaSalle County Station, 
Unit 2, LaSalle County, Illinois

    Date of application for amendment: February 28, 2000, as 
supplemented on April 28, 2000
    Brief description of amendment: The amendment increases the 
Technical Specification safety limit for the Minimum Critical Power 
Ratio from 1.08 for two loop operation and 1.09 for single loop 
operation to 1.11 and 1.12, respectively.
    Date of issuance: May 17, 2000.
    Effective date: Immediately, to be implemented within 60 days.
    Amendment No.: 126.
    Facility Operating License No. NPF-18: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 22, 2000 (65 FR 
15377). The April 28, 2000, submittal provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 17, 2000.
    No significant hazards consideration comments received: No.

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 application for amendments: August 4, 1999, as supplemented 
by letter dated April 19, 2000.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) Limiting Conditions for Operation for 
Reactor Coolant System (RCS) Subcooling Margin Monitor in TS Table 
3.3.3-1 and revise the functions associated with surveillance 
requirements for RCS Loops-Test Exceptions in TS 3.4.17. By letter 
dated April 19, 2000, the licensee withdrew the proposal to relocate 
the Auxiliary Feedwater Loss of Offsite Power function from TS 3.3.2-1 
to TS 3.3.2-1. The other changes requested by August 4,1999, 
application were addressed under separate correspondence.
    Date of issuance: May 19, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 186 and 179.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 8, 1999 (64 
FR 48861)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 19, 2000.
    No significant hazards consideration comments received: No

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: August 18, 1999, as supplemented by 
letter dated April 20, 2000.
    Brief description of amendment: The requested change would revise 
Technical Specification 3.5.3, ``Safety Feature Actuation System 
Setpoints,'' and its associated Bases to allow for an increase to the 
low reactor coolant system pressure setpoint. This setpoint change was 
requested to account for additional instrument uncertainties associated 
with cable insulation resistance effects and to allow for the plugging 
of up to 1200 tubes in each steam generator.
    Date of issuance: May 10, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 207.
    Facility Operating License No. DPR-51: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4270). The April 20, 2000, letter provided clarifying information that 
did not change the scope of the August 18, 1999, application and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 10, 2000.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: July 15, 1999, as supplemented by 
letters dated March 29, 2000, April 13, 2000, April 25, 2000, and May 
9, 2000.
    Brief description of amendment: The amendment changed the Technical 
Specifications to institute a Technical Specification Bases Control 
Program and to provide for record retention as specified in the Quality 
Assurance Program Manual.
    Date of issuance: May 9, 2000.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 161
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4274). The supplements dated March 29, 2000, April 13, 2000, April 25, 
2000, and May 9, 2000, did not change the scope of the initial proposed 
no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 9, 2000. No significant hazards 
consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: July 19, 1999.
    Brief description of amendment: The proposed change modifies 
Technical Specification 4.5.2.f.2 by increasing the performance 
requirement for the low pressure safety injection pumps.
    Date of issuance: May 10, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No. 162.

[[Page 34754]]

    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4277).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 10, 2000.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric 
Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: July 29, 1999.
    Brief description of amendment: The amendment changed the Technical 
Specifications (TS) to extend the allowable outage time to seven days 
for one containment spray system (CSS) train inoperable. A new ACTION 
has been added to provide a shutdown requirement for the inoperability 
of two CSSs. The associated changes to TS Bases are included. However, 
the licensee requested MODE 4 end state for TS 3.6.2.1 is being 
deferred.
    Date of issuance: May 15, 2000.
    Effective date: As of the date of issuance and shall be implemented 
90 days from the date of issuance.
    Amendment No.: 163.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 9, 2000 (65 FR 
6406). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 15, 2000.
    No significant hazards consideration comments received: No.

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating 
Station (LGS), Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: June 22, 1999, as supplemented 
January 3, 2000.
    Brief description of amendments: The amendments remove the 
recirculation system motor generator set stop surveillance requirement 
from the LGS Units 1 and 2 Technical Specifications.
    Date of issuance: May 8, 2000.
    Effective date: Both units--As of date of issuance, to be 
implemented within 30 days.
    Amendment Nos.: 142 and 104.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 8, 1999 (64 
FR 48864). The January 3, 2000, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination or expand the scope of the original Federal 
Register Notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 8, 2000.
    No significant hazards consideration comments received: No

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing

(Exigent Public Announcement or Emergency Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room).
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By June 14, 2000, 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

[[Page 34755]]

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 electronically from the ADAMS Public Library 
component on the NRC Web site, http://www.nrc.gov (the Electronic 
Reading Room). If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or an Atomic 
Safety and Licensing Board, designated by the Commission or by the 
Chairman of the Atomic Safety and Licensing Board Panel, will rule on 
the request and/or petition; and the Secretary or the designated Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: May 8, 2000.
    Description of amendment request: The amendment revises Technical 
Specification Surveillance Requirement 3.8.1.9 to increase the limit 
for the peak transient voltage measured following a full-load rejection 
by the emergency diesel generator that is being tested.
    Date of issuance: May 9, 2000.
    Effective date: As of its date of issuance and shall be implemented 
within 2 days.
    Amendment No.: 140.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: No. The Commission's related 
evaluation of the amendment, finding of emergency circumstances, and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated May 9, 2000.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan, 48226.
    NRC Section Chief: Claudia M. Craig.

    Dated at Rockville, Maryland, this 24th day of May 2000.

For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-13518 Filed 5-30-00; 8:45 am]
BILLING CODE 7590-01-P