[Federal Register Volume 65, Number 183 (Wednesday, September 20, 2000)]
[Notices]
[Pages 56946-56964]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-24021]


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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.

[[Page 56947]]

    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from August 28, 2000, through September 8, 2000. 
The last biweekly notice was published on September 6, 2000 (65 FR 
54083).

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's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC through September 22, 2000. The NRC is 
relocating its Public Document Room to the NRC's headquarters building. 
Effective September 26, 2000, documents may be examined at the NRC's 
Public Document Room, located at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland 20852. The filing of requests 
for a hearing and petitions for leave to intervene is discussed below.
    By October 20, 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 through September 22, 2000 
or at One White Flint North, 11555 Rockville Pike (first floor), 
Rockville, Maryland 20852 effective September 26, 2000, 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.
    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.

[[Page 56948]]

    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 
through September 22, 2000 or at One White Flint North, 11555 Rockville 
Pike (first floor), Rockville, Maryland 20852 effective September 26, 
2000, 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 through September 22, 2000 or at One White 
Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852 effective September 26, 2000, 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 amendments request: July 27, 2000.
    Description of amendments request: The proposed license amendments 
would revise Technical Specification (TS) 3.8.5, ``DC Sources--
Shutdown.'' The operability requirements for the DC sources, during 
shutdown conditions, would be revised to require one of the unit's DC 
electrical power subsystems to be operable when in Modes 4 and 5 and 
during movement of irradiated fuel assemblies in the secondary 
containment.
    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 support of this determination, an evaluation of each of the 
three (3) standards set forth in 10 CFR 50.92 is provided below.
    1. Revising the operability requirements for the DC sources, 
during shutdown conditions, to require one of the unit's DC 
electrical power subsystem to be operable when in Modes 4 and 5 and 
during movement of irradiated fuel assemblies in the secondary 
containment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The operability of the DC electrical power sources during Modes 
4 and 5 and during movement of irradiated fuel assemblies in the 
secondary containment ensures that:
    a. The facility can be maintained in the shutdown or refueling 
condition for extended periods;
    b. Sufficient instrumentation and control capability is 
available for monitoring and maintaining the unit status; and
    c. Adequate DC electrical power is provided to mitigate events 
postulated during shutdown, such as an inadvertent draindown of the 
vessel or a fuel handling accident.
    As stated in TSTF-204, Revision 3, worst case design basis 
accidents which are analyzed for operating modes are not as 
significant of a concern during shutdown modes due to lower energy 
levels. The TSs, therefore, require a lesser complement of 
electrical equipment to be available during shutdown than is 
required during operating modes. Specifically, assuming a single 
failure concurrent with a loss of all offsite or all onsite power is 
not required. This concept is consistent with the BSEP TSs, prior to 
conversion to ITS [Improved TS], in that TS 3.8.2.4.2 required 
either Division I or Division II of the DC power distribution system 
to be operable when in Modes 4 and 5 and during movement of 
irradiated fuel assemblies in the secondary containment. The 
operability requirements of the DC electrical power sources for a 
unit in Modes 1, 2, and 3 are not affected by the proposed 
amendments.
    Therefore, the proposed amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Revising the operability requirements for the DC sources, 
during shutdown conditions, to require one of the unit's DC 
electrical power subsystem to be operable when in Modes 4 and 5 and 
during movement of irradiated fuel assemblies in the secondary 
containment will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    Revising the operability requirements of TS 3.8.5 does not 
involve physical modification to the plant and does not introduce a 
new mode of operation. Therefore, there is no possibility of an 
accident of a new or different type.
    3. Revising the operability requirements for the DC sources, 
during shutdown conditions, to require one of the unit's DC 
electrical power subsystem to be operable when in Modes 4 and 5 and 
during movement of irradiated fuel assemblies in the secondary 
containment does not involve a significant reduction in a margin of 
safety.
    The proposed change revises LCO [Limiting Condition for 
Operation] 3.8.5 to require one of the unit's DC electrical power 
subsystems to be operable when the unit is in Modes 4 and 5 and 
during movement of irradiated fuel assemblies in the secondary 
containment. This is acceptable due to the lower energy levels 
involved with potential accidents occurring during shutdown modes 
and because assuming a single failure concurrent with a loss of all 
offsite or all onsite power during such events is not required. This 
is consistent with the TS requirements, as they existed prior to 
conversion to ITS and TSTF-204, Revision 3 which was approved by the 
NRC on February 16, 2000. The operability requirements of the DC 
electrical power sources for a unit in Modes 1, 2, and 3 are not 
affected by the proposed amendments.
    Based on the above, the proposed amendments do not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Richard P. Correia.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: August 22, 2000.
    Description of amendment request: The proposed amendment would 
change: (1) Technical Specification (TS) 3.10.4, ``Rod Insertion 
Limits,'' to allow on-line calibration of the rod position indicator 
(RPI) channels during operating cycle 15, and (2) TS 3.10.6, 
``Inoperable Rod Position Indicator Channels,'' to allow extended RPI 
deviation limits during cycle 15.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 56949]]

As required by 10 CFR 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:

    1. Does the proposed license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    No. Neither the probability not the consequences of an accident 
previously analyzed is increased due to the proposed changes. All 
peaking factors will remain within the limits of the Technical 
Specifications. Both the shutdown margin and the axial flux 
difference will be maintained within the limits of the Technical 
Specifications. There will be no fuel damage due to the changes. All 
design and safety criteria will be met. Therefore, the proposed 
changes would not involve a significant increase in the probability 
or in the consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The changes will not create the possibility of a new or 
different kind of accident. The calibration will be performed using 
plant procedures that have been reviewed and approved by Con 
Edison's Station Nuclear Safety Committee (SNSC). It has been shown 
that even with the new RPI deviation bands and on-line calibration, 
all power distribution limits will be met. Therefore, the proposed 
changes would not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed amendment does not involve a significnat 
reduction in the margin of safety. There will be no change in the 
power distribution limits used in the design and safety analyses and 
the required shutdown margin will be maintained. It has been shown 
that there is no fuel failure as a result of this change. Therefore, 
the proposed changes do not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Section Chief: Marsha Gamberoni.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of amendment request: August 28, 2000.
    Description of amendment request: The proposed amendment would 
revise the date for implementation of the Palisades Plant Improved 
Technical Specifications (ITS) from on or before October 31, 2000, to 
on or before December 31, 2000. The current implementation date was 
established by previous Amendment No. 189, dated November 30, 1999, in 
which the NRC staff stated that Amendment No. 189 was ``effective as of 
its date of issuance and shall be implemented on or before October 31, 
2000.'' The proposed amendment would change this date to December 31, 
2000.
    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 discussion of these [10 CFR 50.92] standards as they relate to 
this request follows to show that operation of the facility in 
accordance with the proposed change does not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change is administrative in nature in that it 
simply extends the date for implementation of the Improved Technical 
Specifications (ITS) from October 31, 2000 to December 31, 2000. The 
proposed extension of the ITS implementation date is necessary in 
order to allow for additional Operations shift crew training and 
readiness assessment, as well as a longer transition period of 
operating the plant using the Current Technical Specifications (CTS) 
and ITS in parallel. These actions are considered essential to 
proper ITS implementation. Until ITS are implemented, the previously 
approved CTS will remain in effect and the unit will continue to be 
operated in accordance with the NRC approved CTS requirements.
    The proposed change is administrative in nature, and does not 
involve any changes to the design or operation of the Palisades 
Plant which may affect the probability or consequences of an 
accident previously evaluated in the Updated Final Safety Analysis 
Report (UFSAR). Previously evaluated accident precursors or 
initiators are not affected and, as a result, the probability of 
accident initiation will remain as previously evaluated. There will 
be no impact on the capability of any structures, systems or 
components to perform their credited safety functions to prevent an 
accident or mitigate the consequences of an accident previously 
evaluated. Therefore, the probability or consequences of a 
postulated accident previously evaluated in the UFSAR are not 
increased as a result of the proposed change.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Extension of the date for ITS implementation is an 
administrative change. The proposed change does not involve any 
changes to the physical structures, components, or systems of the 
Palisades Plant. Since the change is administrative, there will be 
no impact on the process variables, characteristics, or functional 
performance of any structures, systems or components in a manner 
that could create a new failure mode. Further, the change will not 
introduce any new modes of plant operation or eliminate any actions 
required to prevent or mitigate accidents. Thus, operation in 
accordance with the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    (3) Involve a significant reduction in a margin of safety.
    Extension of the date for ITS implementation is an 
administrative change. The proposed change does not involve any 
hardware changes or physical alteration of the plant and the change 
will have no impact on the design, design basis, or operation of the 
plant. The change will not eliminate any requirements, impose any 
new requirements, or alter any physical parameters which could 
reduce any margin of safety. The continued operation of Palisades in 
accordance with the previously approved Current Technical 
Specifications assures the proposed change will not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Arunas T. Udrys, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Section Chief: Claudia M. Craig.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of amendment request: July 18, 2000, supplemented by letter 
dated August 22, 2000.
    Description of amendment request: The proposed amendments would 
move Technical Specifications Sections 3.3.11, 3.3.12, and 3.3.13 (and 
the corresponding Bases) that specify the Main Steam Line Break 
requirements by renumbering them 3.3.25, 3.3.26, and 3.3.27 and 
indicating that these requirements will remain in effect for each unit 
until after the Automatic Feedwater Isolation System (AFIS) is 
installed on the unit. In addition, the proposed amendments would 
incorporate requirements and Bases for a new AFIS that will become 
effective when the modification is installed on each unit. These 
requirements will

[[Page 56950]]

become Sections 3.3.11, 3.3.12, and 3.3.13.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    No. There is no significant increase in the probability of a 
loss of Main Feedwater (MFW) or Emergency Feedwater (EFW) event due 
to spurious actuation of the Automatic Feedwater Isolation System 
(AFIS). AFIS provides a means of automatic response to improve the 
ability to isolate MFW and EFW to mitigate containment 
overpressurization and steam generator tube stresses resulting from 
Main Steam Line Break (MSLB) accidents. AFIS replaces the need for 
manual operator actions currently required by emergency operating 
procedures, but these remain as a defense-in-depth. AFIS is highly 
reliable, being designed with two independent trains of diverse 
digital control systems, each having four channels of inputs. The 
AFIS modification will also upgrade some existing components that 
were actuated by MSLB Detection and Feedwater (FDW) Isolation 
Circuitry that were not safety grade to safety grade quality thereby 
improving system reliability. Therefore, the installation of AFIS 
does not involve a 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 kind of accident previously evaluated:
    No. AFIS replaces the MSLB Detection and FDW Isolation Circuitry 
as described in the Updated Final Safety Analysis Report. AFIS is 
capable of determining which steam generator has been affected and 
will isolate MFW and EFW to that steam generator. In this regard, 
AFIS performs the same functions that are currently performed by the 
combination of the MSLB Detection and FDW Isolation Circuitry plus 
the additional operator actions needed to isolate the affected steam 
generator. Safety features have been designed into AFIS to prevent 
spurious actuation. The system must be energized to trip therefore 
AFIS will not cause a trip on loss of power. There are no postulated 
failures such as loss of power that differ from those assumed for an 
analog control system that would prevent proper system actuation. 
The design of the two out of four input logic provides redundancy 
against the affects of single failures that could cause spurious 
actuation. In the unlikely event of spurious actuation, manual 
manipulation of EFW pump controls will override the AFIS trip 
signals. Therefore, AFIS does not introduce hardware failures that 
inhibit proper operation of MFW or EFW. In conclusion, AFIS does not 
create the possibility of a new or different kind of accident from 
any kind previously evaluated.
    (3) Involve a significant reduction in a margin of safety.
    No. The proposed change does not adversely affect any plant 
safety limits, setpoints, or design parameters. The change also does 
not adversely affect the fuel, fuel cladding, Reactor Coolant 
System, or containment integrity. For a postulated feedwater line 
break (FLB)/MSLB inside containment, AFIS will improve the margin of 
safety by reducing the mass and energy release to containment. AFIS 
will also improve the margin of safety for departure from nucleate 
boiling by minimizing the overcooling transient from a FLB/MSLB. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: Richard L. Emch, Jr.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of amendment request: August 10, 2000.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3/4.9.4, ``Refueling Operations, 
Containment Building Penetrations,'' by deleting the requirements for 
the containment purge and exhaust system and by revising the closure 
requirements for containment building penetrations to require that 
containment penetrations are capable of being closed during the 
handling of irradiated fuel within the containment.
    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:

Criterion 1--Does Not Involve a Significant Increase in the Probability 
or Consequences of an Accident Previously Evaluated

    The containment purge and exhaust system is not considered an 
accident initiator nor do the proposed changes result in any 
physical change to the plant design. Therefore the probability of an 
accident previously analyzed remains unchanged. In addition, the 
containment purge and exhaust system filtration units are not 
credited in the ANO-2 [Arkansas Nuclear One, Unit 2] safety analysis 
in limiting offsite dose consequences during an accident. 
Furthermore, the system is designed to automatically isolate, as 
required by ANO-2 TS [Technical Specification] 3.3.3.1, Table 3.3-6 
upon receipt of a high radiation signal when in operation in Modes 5 
and 6. Since the containment purge and exhaust system is credited 
only for long-term post accident cleanup efforts and will continue 
to be tested to ensure the filtration system remains effective in 
supporting such efforts, the consequences of an accident previously 
evaluated remains unchanged.
    The opening of a containment penetration during the handling of 
irradiated fuel within the containment building is limited to Mode 6 
with the core flooded to refueling level 
( 23 feet of borated water above the fuel) by the 
applicability of TS 3.9.4. Such openings are strictly controlled by 
safe shutdown programs such as the ANO-2 Shutdown Operations 
Protection Plan (SOPP) and the Outage Risk Management Guidelines 
(ORMG). A containment penetration being open during the handling of 
irradiated fuel does not result in an increase in the probability of 
an accident that has been previously evaluated.
    ANO [Arkansas Nuclear One] submitted the radiological dose 
consequences of a fuel handling accident within the containment 
building to the NRC [Nuclear Regulatory Commission], illustrating 
that without a containment building, the offsite dose consequences 
due to a fuel handling accident inside containment would remain well 
within 10 CFR 100 limits. This evaluation was approved by the NRC in 
Amendment 166 to the ANO-2 Operating License and referenced in the 
aforementioned Amendment 203 to the ANO-2 Operating License in 
support of allowing the equipment hatch and/or personnel air locks 
to remain open during fuel handling activities. Since the above 
evaluation assumes no credit for ``containment'' and subsequently 
illustrates that the resulting offsite dose consequences are 
acceptable, the consequences of an accident previously evaluated are 
not adversely impacted.
    The proposed revision of penetration closure methods does not 
impact any accident previously analyzed or impact the consequences 
of such an accident. The licensee will continue to be accountable 
for ensuring adequate and timely closure of each containment 
penetration should such closure become necessary. Revising the 
examples given in the TSs for establishing closure is, therefore, 
considered risk-neutral and is consistent with the Revised Standard 
Technical Specifications (RSTS) of NUREG-1432.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.

Criterion 2--Does Not Create the Possibility of a New or Different Kind 
of Accident From Any Previously Evaluated

    The containment purge and exhaust system filtration units are 
not credited in the ANO-2 safety analysis in limiting offsite dose 
consequences during an accident. Furthermore, the system is designed 
to automatically isolate, as required by ANO-2 TS 3.3.3.1, Table 
3.3-6, upon receipt of a high radiation signal when in operation in 
Modes 5 and 6. Since the containment purge

[[Page 56951]]

and exhaust system is credited only for long-term post accident 
cleanup efforts and will continue to be tested to ensure the 
filtration system remains effective in supporting such efforts, the 
possibility of a new or different kind of accident being created 
from that previously evaluated remains unchanged.
    The fuel handling accident has previously been addressed in the 
ANO-2 safety analysis. In addition, the offsite dose consequences of 
the fuel handling accident have been found to be acceptable while 
assuming no credit for containment. Therefore, the provision to 
allow penetrations to be opened during the handling of irradiated 
fuel within the containment building does not create the possibility 
of a new or different kind of accident from any previously 
evaluated. The proposed revision of penetration closure methods is 
also not considered an accident initiator. As an added measure of 
safety, however, the appropriate administrative controls required by 
Amendment 203 to the ANO-2 Operating License will be applicable to 
the containment penetrations impacted by the relevant proposals of 
this submittal.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin of 
Safety

    The containment purge and exhaust system is not presently 
permitted to be placed in operation in Modes 1, 2, 3, or 4 and thus 
eliminates one possible path for radiological release to the public. 
The automatic actuations discussed above that act to isolate the 
system during emergency events in Modes 5 and 6 also provide 
assurance that a radiological release will not occur via the 
containment purge and exhaust system flow paths. Furthermore, the 
containment purge and exhaust system filtration units are not 
credited in the ANO-2 safety analysis in limiting offsite dose 
consequences during an accident. Since the containment purge and 
exhaust system is credited only for long-term post accident cleanup 
efforts and will continue to be tested to ensure the filtration 
system remains effective in supporting such efforts, the margin to 
safety remains unchanged.
    ANO-2 has provided sufficient information to illustrate that the 
offsite dose consequences, as a result of a fuel handling accident, 
remain well within 10 CFR [Part] 100 limits, while assuming no 
credit for containment for release mitigation. Since no increase in 
the offsite dose potential is evident due to the opening of 
containment penetrations, the margin to safety is not adversely 
affected by this proposed revision.
    The proposed revision of penetration closure methods does not 
impact the margin to safety. The licensee will continue to be 
accountable for ensuring adequate and timely closure of each 
containment penetration should such closure become necessary.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Dade County, Florida

    Date of amendment request: August 18, 2000.
    Description of amendment request: The proposed amendments would 
revise the current requirements of Technical Specifications (TS) 3.3.2, 
Engineered Safety Features Actuation System Instrumentation, Table 3.3-
2, Items 7.b and 7.c. Specifically, the proposed license amendments 
revise ACTION statement 18 to allow operation of the units with both 
channels of undervoltage protection bypassed for up to 8 hours to allow 
performance of the monthly surveillance without placing the units in a 
condition prohibited by the TS. In addition, an administrative change 
to Item 7.b. of TS Tables 3.3-2, 3.3-3, and 4.3-2 is requested to 
change ``Degraded Voltage'' to ``Undervoltage'' to make it consistent 
with the Updated Final Safety Analysis Report description.
    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.
    Approval and implementation of this amendment will have no 
effect on the probability or consequences of accident previously 
evaluated. The proposed changes allow performance of the required 
surveillance without placing the plant in a condition prohibited by 
the Technical Specifications. The undervoltage and degraded voltage 
protection schemes of the 480 volt load centers are not affected. 
Therefore, there will be no impact on any accident probabilities by 
the approval of this amendment. Therefore, the proposed amendments 
do not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    (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 previously evaluated.
    The proposed changes do not alter the design, physical 
configuration, or modes of operation of the plant. No changes are 
being made to the plant that would introduce any new accident causal 
mechanisms. The proposed Technical Specification changes do not 
impact any plant systems that are accident initiators, since the 480 
volt undervoltage and degraded voltage protection logics are not 
affected. The proposed change allows performance of the required 
surveillance without placing the plant in a condition prohibited by 
the Technical Specifications. No new accident causal mechanisms are 
created as a result of NRC approval of the proposed amendments 
request. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed changes do not change the operation, function or 
modes of plant or equipment operation. The proposed changes do not 
change the undervoltage and degraded voltage protection logics of 
the 480 volt load centers. The ability of the 480 volt load center 
voltage protection schemes to detect degraded voltage and initiate a 
signal to the sequencers is maintained. No new hazards or failure 
modes are created or postulated which may cause an accident 
different from any accident previously analyzed. The proposed 
changes revise ACTION statement 18 to allow performance of the 
technical specification required surveillances without placing the 
plant in a condition prohibited by the Technical Specifications. 
Therefore, operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    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.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment requests: August 18, 2000.
    Description of amendment requests: The proposed amendments would 
change Technical Specification (TS) 3/4.7.4, ``Essential Service Water 
[ESW] System,'' and the associated Bases to add requirements that would 
support cross-connection to the opposite unit. The proposed amendment 
would also

[[Page 56952]]

delete a provision for a 60-day allowed outage time when an ESW 
flowpath is not available to support the opposite unit's shutdown 
functions. Administrative and editorial changes are also made to 
provide consistency between units, correct typographical errors, 
improve readability, and improve page layout.
    The licensee is submitting this request in accordance with Nuclear 
Regulatory Commission (NRC) Administrative Letter 98-10, 
``Dispositioning of Technical Specifications that are Insufficient to 
Assure Plant Safety,'' because the current TS requirements are 
nonconservative. The licensee determined that an operable ESW pump may 
be adversely affected by inoperability of an opposite unit ESW pump 
sharing the same header. With open crosstie valves on the header, an 
inoperable pump can permit flow to be diverted from the operable ESW 
pump to the loads on the opposite unit. This could be safety 
significant when the operable pump is supplying accident loads.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    The accidents previously evaluated in Chapter 14 of the Updated 
Final Safety Analysis Report (UFSAR) that are affected by operation 
of the essential service water system are:

1. Loss-of-Coolant Accident (LOCA),
2. Main Steam Line Break,
3. Feedwater Line Break,
4. Loss of Feedwater,
5. Combinations of the above accidents with loss of offsite power,
6. Appendix R fire, and
7. Flooding.

    The closing of an ESW unit crosstie valve to isolate an 
operating ESW pump from an inoperable loop will not increase the 
probability of occurrence of the affected accidents. Closing these 
valves will not affect the initiators of any previously analyzed 
accidents. It prevents flow in an operating loop from being reduced 
below design basis by flow diversion to an inoperable loop. This 
action will not affect the initiating frequency of any LOCAs, main 
steam line or feedwater line breaks, or loss of feedwater events, 
nor will it cause or increase the frequency of an Appendix R fire.
    With respect to flooding, closing the ESW unit crosstie valves 
may reduce the extent of flooding should a break occur in the ESW 
system. It does not contribute to the probability of an ESW system 
pipe break occurring. Therefore, closing the unit crosstie valve(s) 
as directed by the revised T/S requirements is a conservative change 
relative to flooding.
    Closing an ESW unit crosstie valve to isolate an operating ESW 
pump from an inoperable loop will not increase the consequences of 
any accident previously evaluated in the [Updated Final Safety 
Analysis Report] UFSAR. This configuration does not prevent the ESW 
system from meeting its design basis flow requirements because these 
flows are set with the crosstie valves closed.
    As long as the ESW design basis flow requirements are met, this 
proposed change is bounded by the current analysis of record with 
respect to consequences. No new release paths are created and the 
frequency of release is not increased by closing the unit crosstie 
valve when required by the revised requirements. Preventing the 
diversion of flow from an operating to an inoperable loop will 
reduce the probability of equipment malfunction that could lead to 
an increase in the consequences of an accident. Loss of offsite 
power in conjunction with any of the affected accidents will not be 
impacted by closure of the crosstie valve(s) because the valves 
receive emergency power.
    The change to delete the additional 60-day allowed outage time 
(AOT) of the shutdown flowpath to the opposite unit is a 
conservative change that only increases the availability of the 
shutdown flowpath.
    The change to add T/S 4.0.5 to the Unit 2 surveillance is a 
conservative change that corrects an editorial oversight. 
Surveillance testing under T/S 4.0.5 has been previously evaluated 
and approved.
    The remaining changes are administrative in nature and are not 
intended to change the meaning of the T/S or associated Bases.
    Therefore, these changes cannot increase the consequences or 
probability of occurrence of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Closing an ESW unit crosstie valve to isolate an operating ESW 
pump from an inoperable loop will not create the possibility of an 
accident of a new or different type than any previously evaluated. 
Operation with closed crosstie valves is not the normal operating 
lineup but it is not precluded and applicable procedures recognize 
they may be closed. Therefore, no system/component interfaces are 
affected, nor are new ones created that would contribute toward a 
new or different accident. As described in question 1 above, 
operation with closed crosstie valves is bounded by the current 
analysis for affected accidents, even if these are combined with a 
loss of offsite power. Other single failures in conjunction with 
this change, such as the loss of one train of emergency diesel 
generators on one unit, will not create an accident that is not 
bounded by the current analysis of record.
    The change to delete the additional 60-day AOT of the shutdown 
flowpath to the opposite unit is a conservative change that only 
increases the availability of the shutdown flowpath.
    The change to add T/S 4.0.5 to the Unit 2 surveillance is a 
conservative change that corrects an editorial oversight. 
Surveillance testing under T/S 4.0.5 has been previously evaluated 
and approved and is included in the Unit 1 surveillance 
requirements.
    The remaining changes are administrative/editorial in nature and 
are not intended to change the meaning of the T/S or associated 
Bases.
    Therefore, this proposed change does not increase the 
possibility of a new or different kind of accident than previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Closing an ESW unit crosstie valve to isolate an operating ESW 
pump from an inoperable loop ensures the single-failure design of 
the ESW system will be maintained. In this manner, the system will 
continue to perform its required function and ensure that margins of 
safety is maintained.
    The change to delete the additional 60-day AOT of the shutdown 
flowpath to the opposite unit is a conservative change that assures 
the availability of the shutdown flowpath.
    The change to add T/S 4.0.5 to the Unit 2 surveillance is a 
conservative change that corrects an editorial oversight. 
Surveillance testing under T/S 4.0.5 has been previously evaluated 
and approved and is included in the Unit 1 surveillance 
requirements.
    The remaining changes are administrative in nature and are not 
intended to change the meaning of the T/S or associated Bases.
    Therefore, these changes do not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: Claudia M. Craig.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment requests: September 1, 2000.
    Description of amendment requests: The proposed amendments would 
clarify Technical Specification (TS) 
3/4.4.4, ``Pressurizer,'' to reflect the current power supply to the 
pressurizer heaters and require two operable trains of pressurizer 
heaters during Modes 1, 2, and 3. The proposed amendments also revise 
the Bases for TS 3/4.4.4 to reflect these changes and to clarify the 
purpose of the pressurizer heaters.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 56953]]

issue of no significant hazards consideration, which is presented 
below:

    2. Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    The proposed changes do not affect any accident initiators or 
precursors. Neither the pressurizer heaters nor the surveillance 
test is an accident initiator. Therefore, the proposed changes will 
not affect the probability of an accident.
    The pressurizer heaters are not credited to mitigate the 
consequences of any accidents evaluated in the Updated Final Safety 
Analysis Report; however, they are needed during a loss of offsite 
power to provide adequate subcooling margin in the reactor coolant 
system so that natural circulation conditions can be maintained at 
hot standby. The proposed change to reflect the current power supply 
to the pressurizer heaters modifies the surveillance requirement to 
reflect the design and eliminates redundancy with other surveillance 
requirements. Components will continue to be tested just as 
frequently. The proposed change to require two trains of heaters 
instead of one will provide better assurance that the required 
capacity is operable.
    Overall, testing the same components and requiring redundant 
capacity provides assurance that the required function will be 
performed as assumed. As such, the consequences of an accident will 
not significantly increase.
    Therefore, the probability of occurrence or consequences of 
accidents previously evaluated are not increased.
    (2) Does the change create the possibility of a new or different 
kind of accident from any accident previously analyzed?
    The proposed changes do not modify any equipment or the 
operational limits of any equipment. The proposed changes do not 
introduce any new failure mechanisms to the pressurizer or any other 
plant systems. The proposed changes do not change the method by 
which any plant system performs its function.
    Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously analyzed.
    (3) Does the change involve a significant reduction in a margin 
of safety?
    T/S 3.4.4 requires pressurizer heater capacity of at least 150 
kW to provide adequate subcooling margin in the reactor coolant 
system during a loss of offsite power condition to maintain natural 
circulation conditions at hot standby. The proposed changes will 
increase the requirements for pressurizer heaters by requiring two 
trains of pressurizer heaters to be operable with at least 150 kW in 
each train instead of 150 kW total capacity. This provides assurance 
that the required function will be performed as assumed.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: Claudia M. Craig.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment requests: September 1, 2000.
    Description of amendment requests: The proposed amendments would 
change Technical Specification (TS) surveillance requirement 4.6.1.2 
and the associated T/S Bases to address exemptions to leakage rate 
testing specified by 10 CFR 50, Appendix J, ``Primary Reactor 
Containment Leakage Testing for Water-Cooled Power Reactors,'' and 
Regulatory Guide 1.163, ``Performance-Based Containment Leak-Test 
Program,'' dated September 1995.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    The proposed in-service testing does not affect accident 
initiators or precursors. The proposed ASME Section XI in-service 
test is routinely performed to collect data while the plant is in 
Mode 3. Conducting the containment leakage test in Mode 3 rather 
than prior to Mode 4 entry does not affect the probability of an 
accident. Excessive containment leakage is not a factor until after 
an accident has already occurred.
    The proposed in-service testing does not affect the containment 
leakage rate limits. Therefore, the consequences of an accident are 
unchanged. The proposed change to conduct the testing in Mode 3 
would not significantly increase the consequences of an accident. In 
order to have a release through the modified closed piping systems, 
there would need to be a loss-of-coolant accident concurrent with a 
through-wall leak, with enough pressure in containment to overcome 
main steam system pressure. These conditions are extremely unlikely 
to occur simultaneously in Modes 3 and 4.
    Therefore, the probability of occurrence or the consequences of 
accidents previously evaluated are not increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not introduce any additional physical 
interface with plant equipment. Therefore, the proposed changes do 
not degrade the reliability of systems, structures, or components or 
create a new accident initiator or precursor. No new failure modes 
are created. The proposed changes demonstrate the leak-tight 
integrity of the affected portions of the containment barrier 
through the performance of a visual inspection for through-wall 
leakage.
    Therefore, the change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    T/S 3.6.1.2 is based on limiting total containment leakage 
volume to the value assumed in the accident analysis at peak 
accident pressure. The proposed change does not change the allowable 
leakage rates.
    Since the in-service test is performed at a significantly higher 
pressure than the Type A test and the in-service test acceptance 
criterion is zero through-wall leakage, versus a nominal amount 
allowed for the Type A test, the margin of safety will not be 
reduced. The proposed change would demonstrate the leak-tight 
integrity of the steam generator and associated piping, as 
components of the containment barrier, in a fashion at least as 
rigorous as the Type A test. If the leakage from containment is 
maintained within the T/S limit, dose rates at the site boundary 
will not be increased.
    Therefore, the proposed activity does not involve a significant 
reduction in a margin of safety.
    In summary, based upon the above evaluation, I&M has concluded 
that the proposed amendment involves no significant hazards 
consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: Claudia M. Craig.

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook 
Nuclear Plant, Unit 2, Berrien County, Michigan

    Date of amendment request: September 1, 2000.
    Description of amendment request: The proposed amendment would 
resolve an unreviewed safety question dealing with the licensee 
revising the Unit 1 and 2 safety analyses to incorporate changes 
regarding modeling of pressurizer heater operation and spray 
effectiveness as they relate to certain transients that are analyzed 
for pressurizer overfill. As a part of the revision to the Unit 2 
analyses only, the licensee proposes to change the value of

[[Page 56954]]

the moderator temperature coefficient (MTC) assumed as an initial 
condition for these transients. The licensee evaluated the proposed 
change to the MTC value pursuant to 10 CFR 50.59 and determined that 
the proposed change constituted an unreviewed safety question. 
Therefore, in accordance with 10 CFR 50.90 the licensee is seeking 
approval on its use of a different value for MTC as an input assumption 
for analyses of these transients on Unit 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    The proposed change (i.e., revise MTC assumption) and changes 
already implemented (i.e., revised modeling of pressurizer heaters 
and sprays) result in more conservative modeling of transient 
analyses and do not involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated. The proposed change and changes already implemented would 
affect the analyzed reactor coolant system (RCS) response to a [loss 
of all non-emergency alternating current] LOAC or [loss of normal 
feedwater] LONF transient. Operational occurrences that are 
postulated to occur on a moderate frequency, such as the LOAC and 
LONF transients, are analyzed to ensure they do not generate a more 
serious plant condition without other faults occurring 
independently. Specifically, these events are analyzed to ensure 
that pressurizer overfill would not occur. If pressurizer overfill 
occurs, liquid could pass through the power-operated relief valves 
or the safety valves. Since these valves are qualified to pass steam 
and not liquid, there is a potential to fail one of these valves in 
the open position, creating an uncontrolled release of primary 
coolant. An uncontrolled release of primary coolant through a 
failed-open relief or safety valve would be considered a small break 
loss-of-coolant accident (SBLOCA). Because a loss-of-coolant 
accident is a more serious condition than a LOAC or LONF transient, 
this would constitute a violation of the acceptance criterion 
discussed above. The changes already implemented affect the approach 
to modeling the pressurizer heaters and sprays in the accident 
analysis and the proposed Unit 2 MTC change affects an input 
assumption to the analyses, but none of these changes result in a 
revision to the acceptance criteria for a change in the probability 
of occurrence of these events.
    The changes to the pressurizer heater and spray modeling 
assumptions result in a more conservative outcome for the LOAC and 
LONF transients. When considered in concert with the proposed change 
to reduce the assumed MTC value, the revised LOAC and LONF analyses 
yield acceptable results (pressurizer overfill does not occur). 
Pressurizer heaters and sprays are control systems that are used to 
modulate the primary coolant pressure during normal operation, and 
during certain postulated accident scenarios. Neither of these 
control systems are considered precursors or initiators to any 
accidents described in the Updated Final Safety Analysis Report 
(UFSAR). The change to conservatively assume the heaters are in 
operation during a LOAC or LONF transient does not affect the actual 
design or operation of the heaters during any mode of operation. 
Similarly, revising the assumed effectiveness of the sprays in the 
LOAC and LONF accident analyses has no effect on the actual design 
or operation of the sprays. Consequently, the changes to the assumed 
operation of the sprays and heaters in the LOAC and LONF accident 
analyses would not cause either of these control systems to become 
an initiator or precursor of an accident. The MTC is an analysis 
input that affects the way the plant responds during a temperature 
transient. Reducing the MTC assumed in the analysis to a more 
restrictive value will result in a less severe response of the 
reactor core and RCS to the LOAC and LONF transients. The MTC 
assumed for a safety analysis does not initiate any accident 
scenarios. Changing the MTC as an assumed input to the analysis does 
not result in an increase in the frequency of any initiating events. 
Therefore, these changes do not increase the probability of a 
previously evaluated accident.
    The operation of pressurizer heaters and sprays has no direct 
impact on radiological consequences of a LOAC, LONF, or any other 
previously analyzed event. Similarly, the assumed MTC does not 
directly impact the source term or radiological release pathways for 
any previously analyzed events. Revising the LOAC and LONF analyses 
to conservatively model the pressurizer heaters and sprays and to 
assume a more restrictive MTC value results in precluding the 
occurrence of a pressurizer overfill condition. Consequently, the 
revised LOAC and LONF analyses demonstrate that these transients 
would not progress to the occurrence of a SBLOCA. Therefore, these 
analytical changes do not result in an increase in the consequences 
for these transients.
    Therefore, the probability of occurrence or the consequences of 
accidents previously evaluated are not significantly increased.
    (2) Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The changes in modeling and assumptions for the LOAC and LONF 
transients do not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The changes to 
the modeling of the pressurizer heaters and sprays are analysis 
assumption changes that result in a more conservative outcome for 
the LOAC and LONF transients.
    Because the changes do not alter the design or operation of the 
pressurizer heaters or sprays, they do not introduce any new 
malfunctions. The changes pertain to the correction of analysis 
assumptions for modeling the pressurizer control features for the 
two UFSAR events that have been evaluated. The only potential 
outcome of the application of the heater and revised spray models 
causing the analyses to exceed the acceptance criteria is another 
event (i.e., SBLOCA) that has been evaluated in the UFSAR. 
Consequently, the changes to the modeling of pressurizer heaters and 
sprays in the LOAC and LONF transients cannot affect or create new 
accident initiators or precursors or create the possibility of a new 
or different kind of accident.
    The MTC is an analysis input that affects the way a plant 
responds during a temperature transient. Reducing the MTC assumed in 
the analysis to a more restrictive value will result in a less 
severe response of the reactor core and RCS to the LOAC and LONF 
transients. As used in these analyses, the assumed MTC value is not 
a factor in initiating any accident scenarios. Consequently, the 
application of a lower MTC value to the analyses of the LOAC and 
LONF transients cannot affect or create new accident initiators or 
precursors or create the possibility of a new or different kind of 
accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    (3) Does the change involve a significant reduction in a margin 
of safety?
    The changes to modeling and assumptions for the LOAC and LONF 
transients do not involve a significant reduction in a margin of 
safety. The changes to the modeling of the pressurizer heaters and 
sprays result in a more conservative outcome for the LOAC and LONF 
transients. The acceptance criterion for events analyzed for 
pressurizer overfill, such as the LOAC and LONF transients, is that 
the pressurizer does not reach a water-solid condition. In order for 
the Unit 2 analyses to meet this acceptance criterion, it was 
necessary to change the assumed MTC from a positive value to zero at 
full-power conditions. However, a positive MTC has been assumed in 
previous NRC analyses of these transients in support of past [Donald 
C. Cook Nuclear Plant] CNP license amendments. Specifically, the 
NRC's approval of the current Unit 2 Technical Specification (T/S) 
3/4.1.1.4 MTC curve (License Amendment 107 to DPR-74) was 
predicated, in part, on the basis that the safety analysis 
assumptions remain valid. Because a positive MTC was assumed in 
previous safety analyses, and the proposed MTC value is less 
conservative than the positive value assumed in those previous 
safety analyses, this activity is a reduction in the margin of 
safety.
    T/S 3/4.1.1.4, Figure 3.1-2, specifies the operational limits 
for the MTC. The T/S allows a constant MTC of +5 pcm/ deg.F for core 
power levels from 0% to 70%. Above 70% power, the allowed MTC value 
ramps down linearly to 0 pcm/ deg.F at full power. The basis for the 
limitations on MTC are provided to ensure that the value of this 
coefficient remains within the limiting conditions assumed in the 
UFSAR accident and transient analyses. Although the revised initial 
MTC value assumed in these analyses is reduced from that assumed in 
the current

[[Page 56955]]

analyses of record in the CNP UFSAR, it is still within the 
requirements of Unit 2 T/S 3.1.1.4 for 100% power. Thus, the revised 
MTC value remains bounding for full-power operation. Furthermore, 
the analyses for the LONF and LOAC scenarios both assume an initial 
reactor power of 102%, which also bounds full power operation. 
Consequently, the revised assumption for the Unit 2 MTC ensures that 
the conditions assumed for the transient evaluation still bound the 
most limiting plant operating conditions and are consistent with the 
requirements in T/S 3.1.1.4. Thus, the basis for approval of the 
current Unit 2 MTC curve, as specified by the safety evaluation 
report that approved this curve (Amendment No. 107 to DPR-74), 
remains valid.
    A sensitivity study was performed to confirm that the basis for 
establishing an MTC of 0 pcm/ deg.F at full power bounds partial-
power conditions with the corresponding positive MTC. Specific LOAC 
and LONF calculations were performed by varying (reducing) the 
nominal core power levels and assuming the corresponding positive 
MTC values at each core power level. The result of the study 
confirmed that, for both the LOAC and LONF events, the full power 
case with an MTC value of 0 pcm/ deg.F bounds the case with a 
positive MTC initialized at a lower power level.
    By revising the assumed MTC value from a positive value to zero, 
the Unit 2 LOAC and LONF analyses demonstrate that the analysis 
acceptance criteria are met (i.e., pressurizer overfill does not 
occur) and bound the positive MTC cases at lower power levels. 
Therefore, the combination of these three analytical modeling 
changes results in an acceptable analytical outcome for Unit 2. 
Furthermore, the zero MTC value is still within the requirements of 
Unit 2 T/S 3.1.1.4 for 100% power. Thus, the revised MTC value 
remains bounding for full-power operation. Although the proposed MTC 
assumed in the LOAC and LONF analyses are a reduction in the margin 
provided to the NRC in previous evaluations of these transients, the 
use of the full-power MTC is consistent with the plant T/S and is 
bounding for full-power operation and partial-power operation at the 
corresponding MTC value allowed by the plant T/S.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: Claudia M. Craig.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point 
Nuclear Station Unit No. 1, Oswego County, New York

    Date of amendment request: November 30, 1999; as supplemented on 
August 15, 2000.
    Description of amendment request: The licensee proposed to amend 
the unit's Technical Specifications (TSs), Section 3.4.4, ``Emergency 
Ventilation System [EVS],'' and Section 3.4.5, ``Control Room Air 
Treatment [CRAT] System,'' to require testing consistent with American 
Society for Testing and Materials (ASTM) Standard D3803-1989. Currently 
Section 3.3.4 specifies the American National Standards Institute 
(ANSI) standard N510-1980. The licensee's application for amendment is 
a response to the NRC's Generic Letter (GL) 99-02, ``Laboratory Testing 
of Nuclear-Grade Activated Charcoal.'' The staff had previously 
published a notice (65 FR 9009, February 23, 2000) for the licensee's 
November 30, 1999, submittal. The licensee's August 15, 2000, submittal 
revises the original submittal by increasing the charcoal bed testing 
efficiency of the EVS from 95 percent to 99.5 percent, and requiring 
the pressure drop across the CRAT System high efficiency particulate 
air (HEPA) filters and charcoal adsorber banks to be demonstrated to be 
less than 1.5 inches of water.
    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 operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed TS change will require the demonstration that the 
pressure drop across the combined HEPA filters and charcoal adsorber 
banks is less than 1.5 inches of water at system design flow rate 
( 10%). The CRAT System does not involve initiators or 
precursors to an accident previously evaluated, as this system 
performs mitigative functions in response to an accident. Failure of 
this system would result in the inability to perform its mitigative 
function, but would not increase the probability of an accident. 
Therefore, the probability of an accident previously evaluated is 
not increased.
    The NMP1 [Nine Mile Point Unit 1] CRAT System is designed to 
limit doses to control room operators to less than the values 
allowed by General Design Criterion 19. This system contains HEPA 
filters and activated charcoal adsorber banks that are required by 
TS to have a combined pressure drop across them of less than 6 
inches of water. The proposed TS change to require a combined 
pressure drop of less than 1.5 inches of water will assure the 
capability of the CRAT System to maintain the required minimum 
positive pressure in the Control Room complex. Therefore, the 
proposed change will not involve a significant increase in the 
consequences of an accident previously evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed TS change will revise the allowable pressure drop 
across the CRAT System HEPA filters and charcoal adsorber banks to 
less than 1.5 inches of water at system design flow rate 
( 10%). This change will not involve placing the system 
in a new configuration or operating the system in a different manner 
that could result in a new or different kind of accident. 
Maintaining a combined pressure drop across the HEPA filters and 
charcoal adsorber banks to less than 1.5 inches of water will assure 
system capability of maintaining the required minimum positive 
pressure in the Control Room complex. Therefore, the proposed change 
will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The proposed TS change will not adversely affect the performance 
characteristics of the CRAT System, nor will it affect the ability 
of the system to perform its intended function. The combined 
pressure drop across the CRAT System HEPA filters and charcoal 
adsorber banks is demonstrated to determine whether sufficient flow 
exists to maintain the minimum positive pressure in the control room 
assumed in the design basis analysis. The proposed TS change will 
require the combined pressure drop across the HEPA filters and 
charcoal adsorber banks to be less than 1.5 inches of water. This 
will assure system capability to maintain the required minimum 
positive pressure in the Control Room complex. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Marsha K. Gamberoni.

[[Page 56956]]

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

    Date of amendment request: August 16, 2000.
    Description of amendment request: The amendment would (1) remove 
``Offgas Treatment System Explosive Gas Mixture Instrumentation,'' 
Specification 3.7, from the Radiological Effluent Technical 
Specifications (RETS) contained in Appendix B and include reference to 
the Offgas Treatment System Explosive Gas Monitoring Program in 
Administrative Section 6 to the Technical Specifications contained in 
Appendix A; (2) replace the position title of Radiological and 
Environmental Services Manager, contained in the Administrative Section 
6 of Appendix A, with radiation protection manager; and (3) revise 
Plant Staff organization requirements contained in Administrative 
Section 6 to require either the Operations Manager or Assistant 
Operations Manager hold a senior reactor operator license.
    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:

    Operation of the FitzPatrick plant in accordance with the 
proposed amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes simplify the RETS and meet Code of Federal 
Regulation requirements as specified in 10 CFR 50.36. Future changes 
to these requirements will be controlled by 10 CFR 50.59. The 
proposed changes are administrative in nature and do not involve any 
modification to any plant equipment or effect plant operation. 
Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any previously 
evaluated accident.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed changes are administrative in nature, do not 
involve any physical alterations to any plant equipment, and cause 
no change in the method by which any safety related system performs 
its function. Therefore, this proposed amendment will not create the 
possibility of a new or differen[t] kind of accident from any 
accident previously evaluated.
    (3) Involve a significant reduction in a margin of safety.
    The proposed changes are administrative in nature, will not 
alter the basic regulatory requirements, and do not affect any 
safety analyses. Therefore, no margin of safety is reduced as a 
result of these changes.
    Based on the above evaluation, the Authority has concluded that 
these changes do not involve a significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: Marsha K. Gamberoni.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of amendment request: August 28, 2000.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) Section 5.5.2 b. ``Primary Coolant 
Sources Outside Containment'' by changing the system leak test 
frequency from a ``refueling cycle'' to ``at least once every 18 
months.'' The proposed change will also allow the provisions of 
Surveillance Requirement (SR) 3.0.2 to apply to TS 5.5.2 b. (SR 3.0.2 
allows a surveillance to be preformed within 1.25 times the interval 
specified in a Frequency.)
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed changes affect programmatic administrative 
controls of the Vogtle Electric Generating Plant (VEGP) Technical 
Specifications (TS) for leak testing systems or portions thereof 
that are outside containment and could contain highly radioactive 
fluids. Only the interval for leak testing is affected by the 
proposed change, and this interval has no impact on the likelihood 
of any of the initiating events assumed for any accident previously 
evaluated. Therefore, the proposed change will not result in a 
significant increase in the probability of any accident previously 
evaluated. Whereas the current TS require testing at refueling cycle 
intervals or less, the proposed change will specify testing at least 
once per 18 months, and the provisions of Surveillance Requirement 
(SR) 3.0.2 will be applicable. Refueling cycle intervals at VEGP are 
nominally 18 months in duration, but they can vary with unplanned 
outages, power reductions, etc. Under the proposed change, leak 
testing will be performed at 18-month intervals, regardless of 
actual refueling cycle length, and if an extension of that interval 
becomes necessary for systems or portions thereof due to scheduling 
considerations, the provisions of SR 3.0.2 will provide the 
necessary flexibility. However, the maximum extension that can be 
applied is 25% of 18 months or four and one-half months. Leak 
testing will continue at regular intervals, and any necessary 
maintenance to minimize leakage will continue to be performed. 
Therefore, the proposed change will not result in a significant 
increase in the consequences of any accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    No. The proposed change affects only the interval at which leak 
test requirements are performed pursuant to TS 5.5.2.b. The proposed 
change does not alter the operation of the plant or any of its 
equipment, introduce any new equipment, or result in any new failure 
mechanisms or limiting single failures. Therefore, there is no 
potential for a new accident and no changes to the way that an 
analyzed accident will progress. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Do the proposed changes result in a significant reduction in 
a margin of safety?
    No. The proposed change affects only the interval at which leak 
test requirements are performed pursuant to TS 5.5.2.b. Under the 
proposed change, leak testing will be performed at 18-month 
intervals, regardless of actual refueling cycle length, and if an 
extension of that interval becomes necessary for systems or portions 
thereof due to scheduling considerations, the provisions of SR 3.0.2 
will provide the necessary flexibility. However, the maximum 
extension that can be applied is 25% of 18 months or four and one-
half months. Leak testing will continue at regular intervals, and 
any necessary maintenance to minimize leakage will continue to be 
performed. The intent of the program is maintained while providing 
the same scheduling flexibility that is already provided for the 
surveillance requirements of Section 3.0 of the TS. Therefore, the 
proposed change will not 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: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.

[[Page 56957]]

    NRC Section Chief: Richard L. Emch, Jr.

Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar 
Nuclear Plant Units 1 and 2, Rhea County, Tennessee

    Date of amendment request: March 10, 2000.
    Description of amendment request: The proposed amendment would 
change the Operating License to Physical Security/Contingency Plan--
Tamper Indicating/Line Supervision Alarms Testing Frequency at Watts 
Bar Nuclear Plant (WBN) Units 1 and 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    There are no safety-related systems, components, or radiological 
waste systems associated with the tamper indicating/line supervision 
alarms. The proposed change to the Physical Security Plan does not 
involve any physical alterations of plant configuration, changes to 
setpoints, or changes to any operating parameters of the security 
system. The proposed change does not increase the frequency of the 
precursors to design basis events or operational transients analyzed 
in the Watts Bar Final Safety Analysis Report. * * * Consequently, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    There are no safety-related systems, components, or radiological 
waste systems associated with the tamper indicating/line supervision 
alarms. The proposed change to extend the testing frequency cannot 
create a Final Safety Analysis Report type accident. * * * 
Consequently, 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.
    Implementation of this activity will not reduce the margin of 
safety in the Technical Specification as there are no Technical 
Specification requirements associated with the physical security 
system. The proposed amendment to the Physical Security Plan does 
not change or reduce the effectiveness of any security/safeguards 
measures currently in place at WBN. * * * Therefore, the proposed 
change does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. The staff has also reviewed the changes to License Condition 
2.E for Watts Bar Unit 1 Operating License, as well as the change to 
the Security Plan. 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.

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

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

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

    Date of amendment request: May 12, 2000.
    Brief description of amendment request: The proposed amendment 
would revise the standard to which the control room ventilation 
charcoal and supplementary leak collection and release system (SLCRS) 
charcoal must be laboratory tested as specified in: BVPS-1 Technical 
Specification (TS) 4.7.7.1.1.c.2 for the control room emergency 
habitability systems; BVPS-1 TS 4.7.8.1.b.3 for the SLCRS; BVPS-2 TS 
4.7.7.1.d for the control room emergency air cleanup and pressurization 
system; and BVPS-2 TS 4.7.8.1.b.3 for the SLCRS. NRC Generic Letter 99-
02, ``Laboratory Testing of Nuclear-Grade Activated Charcoal,'' dated 
June 3, 1999, requested licensees to revise their TS criteria 
associated with laboratory testing of ventilation charcoal to a valid 
test protocol, which included American Society for Testing and 
Materials (ASTM) D3803-1989. This license amendment request revises the 
charcoal laboratory standard to follow ASTM D3803-1989 for each BVPS 
Unit. This license amendment request also: (1) Revises the minimum 
amount of output in kilowatts needed for the control room emergency 
ventilation system heaters at each BVPS unit; (2) revises BVPS-1 SLCRS 
surveillance testing criteria to be consistent with American Nuclear 
Standards Institute/American Society of Mechanical Engineers N510-1980, 
the BVPS-1 control room ventilation testing, and BVPS-2 SLCRS/control 
room ventilation testing; and (3) makes minor typographical corrections 
and editorial changes.
    Date of publication of individual notice in Federal Register: 
August 29, 2000 (65 FR 52449).
    Expiration date of individual notice: September 28, 2000.

FirstEnergy Nuclear Operating Company, et al., Docket No.50-412, Beaver 
Valley Power Station, Unit No. 2, Shippingport, Pennsylvania

    Date of amendment request: May 1, 2000, as supplemented July 21, 
2000.
    Brief description of amendment request: The proposed amendment 
would: (1) Revise Technical Specification (TS) requirements regarding 
the minimum number of radiation monitoring instrumentation channels 
required to be operable during movement of fuel within the containment; 
(2) revise the Modes in which the surveillance specified by Table 4.3-
3, ``Radiation Monitoring Instrumentation Surveillance Requirements,'' 
Item 2.c.ii is required; (3) revise TS 3.9.4, ``Containment Building 
Penetrations,'' to allow both personnel air lock (PAL) doors and other 
containment penetrations to be open during movement of fuel assemblies 
within containment, provided certain conditions are met; (4) revise 
applicability and action statement requirements of TS 3.9.4. to be for 
only during movement of fuel assemblies within containment; (5) revise 
periodicity and applicability of Surveillance Requirement (SR) 4.9.4.1; 
(6) revise SR 4.9.4.2 to verify flow rate of air to the supplemental 
leak collection and release system (SLCRS) rather than verifying the 
flow rate through the system; (7) add two new SRs, 4.9.4.3 and 4.9.4.4, 
for verification and demonstration of SLCRS operability; (8) modify TS 
3/4.9.9 for the containment purge exhaust and

[[Page 56958]]

isolation system to be applicable only during movement of fuel 
assemblies within containment; (9) revise associated TS Bases as well 
as make editorial and format changes; and, (10) revise the BVPS-2 
Updated Final Safety Analysis Report (UFSAR) description of a fuel-
handling accident (FHA) and its radiological consequences. The changes 
to the BVPS-2 UFSAR reflect a revised FHA analysis that the licensee 
performed to evaluate the potential consequences of having containment 
penetrations and/or the PAL open during movement of fuel assemblies 
within containment. These UFSAR revisions include potential exclusion 
area boundary, low population zone, and control room operator doses as 
a result of an FHA.
    Date of publication of individual notice in Federal Register: 
August 23, 2000 (65 FR 51342).
    Expiration date of individual notice: September 22, 2000.

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

    Date of application for amendments: January 20, 2000, as 
supplemented April 3 and July 7, 2000.
    Brief description of amendments: The amendments revised the 
technical specifications to extend the allowable completion times 
associated with restoration of an inoperable emergency diesel 
generator. The amendments also permitted the performance of the 
24-hour endurance run during Modes 1 and 2.
    Date of issuance: September 1, 2000.
    Effective date: Effective upon completion of the plant 
modifications cited in the April 3, 2000, submittal.
    Amendment Nos.: 114 and 108.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 19, 2000 (65 FR 
21035). The April 3 and July 7, 2000, submittals provided additional 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
September 1, 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: April 18, 2000, as supplemented 
by letter dated July 27, 2000.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) 3.7.10, ``Control Room Area Ventilation 
System,'' and TS 3.7.12, ``Auxiliary Building Filtered Ventilation 
Exhaust System,'' to establish actions to be taken for inoperable 
ventilation systems due to a degraded control room pressure boundary or 
emergency core cooling system pump rooms pressure boundary, 
respectively.
    Date of issuance: September 5, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 187/180.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 31, 2000 (65 FR 
34744).
    The supplement dated July 27, 2000, provided additional 
clarifications that did not change the scope of the April 18, 2000, 
application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 5, 2000.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application of amendments: December 16, 1998; supplemented 
January 25, August 5, and October 4, 1999; and March 29 and June 8, 
2000.
    Brief description of amendments: The amendments revised the 
Technical Specifications associated with the High Pressure Injection 
System.
    Date of Issuance: September 6, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 75 days.
    Amendment Nos.: 314, 314, & 314.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1999 (64 
FR 9187).
    The supplements dated August 5 and October 4, 1999; and March 29 
and June 8, 2000, provided clarifying information that did not change 
the scope of the December 16, 1998, or the January 25, 1998, submittals 
and the proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 6, 2000.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of application for amendment: January 27, 2000.
    Brief description of amendment: The amendment revised the Technical 
Specifications by providing actions associated with inoperable control 
room emergency ventilation or cooling systems during movement of 
irradiated fuel during shutdown modes of operation, when the allowed 
outage times associated with these systems are not met.
    Date of issuance: August 28, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 219.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 22, 2000 (65 FR 
15379).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 28, 2000.
    No significant hazards consideration comments received: No .

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of application for amendment: March 8, 2000, as supplemented 
by letters dated June 13 and August 15, 2000.
    Brief description of amendment: The amendment revised Technical 
Specification Definition 1.12 , ``Core Alteration,'' to explicitly 
define core alteration as the movement or manipulation of any fuel, 
sources, or reactivity control components within the reactor vessel 
with the vessel head removed and fuel in the vessel.
    Date of issuance: September 7, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 220.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 5, 2000 (65 FR 
17914).

[[Page 56959]]

    The August 15, 2000, supplement withdrew the exclusion clause, 
``excluding coupling/uncoupling of control element assemblies,'' from 
the proposed definition in the initial application. The June 13 and 
August 15, 2000, supplemental letters provided clarifying information 
that was within the scope of the original FEDERAL REGISTER notice and 
did not change the staff's initial no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 7, 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 letter 
dated March 29, 2000.
    Brief description of amendment: The amendment creates a new 
Technical Specification (TS) for the Main Feedwater Isolation Valves 
(MFIV) Section modeled after the guidelines of TS 3.7.3 in NUREG-1432. 
Additionally, the letter provides for the Nuclear Regulatory Commission 
staff review of an unreviewed safety question regarding the crediting 
of the Reactor Trip Override feature and Auxiliary Feedwater Pump high 
discharge pressure trip as assisting the operation of the MFIV during 
their required safety function, to close on a Main Steam Isolation 
Signal.
    Date of issuance: September 5, 2000.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 167.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000, (65 
FR 4275 ). The March 29, 2000, supplement provided clarifying 
information that did not expand the scope of the original Federal 
Register notice, or 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 September 5, 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: October 29, 1999, and as supplemented by 
letter dated June 29, 2000.
    Brief description of amendment: Entergy Operations, Inc. (licensee) 
has proposed to revise its Updated Final Safety Analysis Report (UFSAR) 
to discuss the probability threshold for when physical protection of 
safety-related components from tornado missiles is required for certain 
components. The proposed changes involve the use of Nuclear Regulatory 
Commission (NRC) approved probability risk methodology to assess the 
need for additional tornado missile protection and demonstrate that the 
probability of damage due to tornado missiles striking safety related 
components is acceptably low.
    Date of issuance: September 7, 2000.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 168.
    Facility Operating License No. NPF-38: The amendment revised the 
UFSAR.
    Date of initial notice in Federal Register: June 14, 2000 (65 FR 
37426).
    The June 29, 2000, supplement provided clarifying information that 
did not expand the scope of the original Federal Register notice, or 
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 September 7, 2000.
    No significant hazards consideration comments received: No

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

    Date of application for amendments: May 1, 2000.
    Brief description of amendments: The amendments revised the Unit 1 
and 2 Technical Specification (TS) 3/4.6.4.2 Surveillance Requirement 
(SR). The change allows performance of hydrogen recombiner functional 
test at containment pressures greater than 13 psia. This is 
accomplished by measuring the flow under normal or current test 
conditions (e.g., atmospheric pressure) and calculating the expected 
system performance under design basis operating conditions. The 
surveillance was revised to verify that the recombiner flow, when 
corrected to the post-accident design conditions, is greater than or 
equal to the required flow. The corresponding design basis temperature 
for post-accident recombiner operation is included in the SR because it 
is required to correct the test flow to the design basis operating 
conditions. In order to support the calculations necessary to confirm 
the recombiner blower performance, the change included the addition of 
an equation and associated discussion to the bases. The equation will 
correct the measured test flow to a corresponding flow at the design 
basis operating pressure and temperature. In addition to the technical 
change described above, SR 4.6.4.2.b.3 was modified by separating the 
criteria for the system blower performance and heater operation into 
separate parts of the same surveillance to improve the presentation of 
the requirements. Format and editorial changes were included as 
necessary to facilitate the revision of the TS text to conform to the 
current TS page format, and addition of text to the bases.
    Date of issuance: September 7, 2000.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 232 and 114.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 14, 2000 (65 FR 
37427).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 7, 2000.
    No significant hazards consideration comments received: No.

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

[[Page 56960]]

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 through September 22, 2000 or at One White 
Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852 effective September 26, 2000, and electronically from the ADAMS 
Public Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

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

    Date of application for amendments: January 20, 2000, as 
supplemented April 3 and July 7, 2000.
    Brief description of amendments: The amendments revised the 
technical specifications to extend the allowable completion times 
associated with restoration of an inoperable emergency diesel 
generator. The amendments also permitted the performance of the 24-hour 
endurance run during Modes 1 and 2.
    Date of issuance: September 1, 2000.
    Effective date: Effective upon completion of the plant 
modifications cited in the April 3, 2000, submittal.
    Amendment Nos.: 114 and 108.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
    The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 19, 2000 (65 FR 
21035). The April 3 and July 7, 2000, submittals provided additional 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
September 1, 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: April 18, 2000, as supplemented 
by letter dated July 27, 2000.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) 3.7.10, ``Control Room Area Ventilation 
System,'' and TS 3.7.12, ``Auxiliary Building Filtered Ventilation 
Exhaust System,'' to establish actions to be taken for inoperable 
ventilation systems due to a degraded control room pressure boundary or 
emergency core cooling system pump rooms pressure boundary, 
respectively.
    Date of issuance: September 5, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 187/180.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 31, 2000 (65 FR 
34744).
    The supplement dated July 27, 2000, provided additional 
clarifications that did not change the scope of the April 18, 2000, 
application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 5, 2000.
    No significant hazards consideration comments received: No.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application of amendments: December 16, 1998; supplemented 
January 25, August 5, and October 4, 1999; and March 29 and June 8, 
2000.
    Brief description of amendments: The amendments revised the 
Technical Specifications associated with the High Pressure Injection 
System.
    Date of Issuance: September 6, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 75 days.
    Amendment Nos.: 314, 314, and 314.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1999 (64 
FR 9187).
    The supplements dated August 5 and October 4, 1999; and March 29 
and June 8, 2000, provided clarifying information that did not change 
the scope of the December 16, 1998, or the January 25, 1998, submittals 
and the proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 6, 2000.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of application for amendment: January 27, 2000.
    Brief description of amendment: The amendment revised the Technical 
Specifications by providing actions associated with inoperable control 
room emergency ventilation or cooling systems during movement of 
irradiated fuel during shutdown modes of operation, when the allowed 
outage times associated with these systems are not met.
    Date of issuance: August 28, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 219.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 22, 2000 (65 FR 
15379).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 28, 2000.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of application for amendment: March 8, 2000, as supplemented 
by letters dated June 13 and August 15, 2000.
    Brief description of amendment: The amendment revised Technical 
Specification Definition 1.12 , ``Core Alteration,'' to explicitly 
define core alteration as the movement or manipulation of any fuel, 
sources, or reactivity control components within the reactor vessel 
with the vessel head removed and fuel in the vessel.
    Date of issuance: September 7, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.

[[Page 56961]]

    Amendment No.: 220.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register April 5, 2000 (65 FR 
17914).
    The August 15, 2000, supplement withdrew the exclusion clause, 
``excluding coupling/uncoupling of control element assemblies,'' from 
the proposed definition in the initial application. The June 13 and 
August 15, 2000, supplemental letters provided clarifying information 
that was within the scope of the original Federal Register notice and 
did not change the staff's initial no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 7, 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 letter 
dated March 29, 2000.
    Brief description of amendment: The amendment creates a new 
Technical Specification (TS) for the Main Feedwater Isolation Valves 
(MFIV) Section modeled after the guidelines of TS 3.7.3 in NUREG-1432. 
Additionally, the letter provides for the Nuclear Regulatory Commission 
staff review of an unreviewed safety question regarding the crediting 
of the Reactor Trip Override feature and Auxiliary Feedwater Pump high 
discharge pressure trip as assisting the operation of the MFIV during 
their required safety function, to close on a Main Steam Isolation 
Signal.
    Date of issuance: September 5, 2000.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 167.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4275). The March 29, 2000, supplement provided clarifying information 
that did not expand the scope of the original Federal Federal Register 
notice, or 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 September 5, 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: October 29, 1999, and as supplemented by 
letter dated June 29, 2000.
    Brief description of amendment: Entergy Operations, Inc. (licensee) 
has proposed to revise its Updated Final Safety Analysis Report (UFSAR) 
to discuss the probability threshold for when physical protection of 
safety-related components from tornado missiles is required for certain 
components. The proposed changes involve the use of Nuclear Regulatory 
Commission (NRC) approved probability risk methodology to assess the 
need for additional tornado missile protection and demonstrate that the 
probability of damage due to tornado missiles striking safety related 
components is acceptably low.
    Date of issuance: September 7, 2000.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 168.
    Facility Operating License No. NPF-38: The amendment revised the 
UFSAR.
    Date of initial notice in Federal Register: June 14, 2000 (65 FR 
37426).
    The June 29, 2000, supplement provided clarifying information that 
did not expand the scope of the original Federal Register notice, or 
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 September 7, 2000.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit 2, Shippingport, Pennsylvania

    Date of application for amendment: June 17, 1999, as supplemented 
September 15, 1999, and February 15, and June 29, 2000.
    Brief description of amendment: The amendment revised Technical 
Specifications (TSs) Section 3.4.9.1 and associated figures to extend 
the applicability of the heatup and cooldown curves from 10 Effective 
Full Power Years (EFPY) to 15 EFPY. The changes included new heatup and 
cooldown curves developed in accordance with the methodology provided 
in Regulatory Guide 1.99, Revision 2, and Code Case N-640. The 
applicability of TS Section 3.4.9.3, Overpressure Protection Systems, 
was also updated to 15 EFPY, and the maximum allowable power-operated 
relief valve setpoints for the over pressure protection system were 
revised. Revisions to the TS Bases were also made.
    Date of issuance: September 6, 2000.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment No: 113.
    Facility Operating License No. NPF-73. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 17, 1999, (64 
FR 62707). The September 15, 1999, and February 15, and June 29, 2000, 
letters provided supplemental and revised information, but did not 
change the initial proposed no significant hazards consideration 
determination or expand the amendment beyond the scope of the initial 
notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 6, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 1, 2000.
    Brief description of amendments: The amendments revised the Unit 1 
and 2 Technical Specification (TS) 3/4.6.4.2 Surveillance Requirement 
(SR). The change allows performance of hydrogen recombiner functional 
test at containment pressures greater than 13 psia. This is 
accomplished by measuring the flow under normal or current test 
conditions (e.g., atmospheric pressure) and calculating the expected 
system performance under design basis operating conditions. The 
surveillance was revised to verify that the recombiner flow, when 
corrected to the post-accident design conditions, is greater than or 
equal to the required flow. The corresponding design basis temperature 
for post-accident recombiner operation is included in the SR because it 
is required to correct the test flow to the design basis operating 
conditions. In order to support the calculations necessary to confirm 
the recombiner blower performance, the change included the addition of 
an equation and associated discussion to the bases. The equation will 
correct the measured test flow to a corresponding

[[Page 56962]]

flow at the design basis operating pressure and temperature. In 
addition to the technical change described above, SR 4.6.4.2.b.3 was 
modified by separating the criteria for the system blower performance 
and heater operation into separate parts of the same surveillance to 
improve the presentation of the requirements. Format and editorial 
changes were included as necessary to facilitate the revision of the TS 
text to conform to the current TS page format, and addition of text to 
the bases.
    Date of issuance: September 7, 2000.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 232 and 114.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 14, 2000 (65 FR 
37427).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 7, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 20, 1999.
    Brief description of amendments: The amendments relocated the 
following Technical Specification (TS) items to the Licensing 
Requirements Manual:

In-core Detectors (Unit 1 and 2),
Chlorine Detection System (Unit 1 and 2)
Turbine Over-speed Protection (Unit 2 only),
Crane Travel Spent Fuel Pool Building (Unit 1 and 2).

    Additionally, certain information on the Remote Shutdown Panel 
Monitoring Instrumentation was moved to the Updated Final Safety 
Analysis Report. Finally, additions to the TS Bases, and certain 
editorial and format changes were made.
    Date of issuance: September 7, 2000.
    Effective date: As of date of issuance and shall be implemented 
within 90 days.
    Amendment Nos.: 233 and 115.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 17, 1999 (64 
FR 62709).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 7, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 17, 1999, as supplemented 
by letters dated January 17, March 1, March 20, May 9, and August 21, 
2000.
    Brief description of amendment: This amendment revised multiple 
surveillance requirements to support a 24-month operating cycle.
    Date of issuance: August 29, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 115.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46438).
    The supplemental information contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 29, 2000.
    No significant hazards consideration comments received: No.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: July 14, 2000.
    Brief description of amendment: The amendment changes the 
implementation dates of the Improved Technical Specifications, 
previously issued by Amendment No. 91, and requirements for the 
Oscillation Power Range Monitor, previously issued by Amendment No. 92, 
from August 31, 2000, to December 31, 2000.
    Date of issuance: August 29, 2000.
    Effective date: As of the date of issuance to be implemented no 
later than December 31, 2000.
    Amendment No.: 94.
    Facility Operating License No. NPF-69: Amendment revised the 
Operating License.
    Date of initial notice in Federal Register: July 27, 2000 (65 FR 
46183).
    The staff's related evaluation of the amendment is contained in a 
Safety Evaluation dated August 29, 2000.
    No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power 
Plant, Unit 3, Humboldt County, California

    Date of application for amendment: December 1, 1999
    Brief description of amendment: The amendment revised the technical 
specifications (TS) to reflect relocation of fire protection 
requirements from the TS to the Defueled Safety Analysis Report, 
quality assurance audit requirements from the TS to the Quality 
Assurance Plan and modification of the administrative controls section 
of the TS to reflect the current facility organization.
    Date of issuance: August 31, 2000.
    Effective date: August 31, 2000, and shall be implemented no later 
than 60 days from the date of issuance. Implementation shall include 
the relocation of technical specification requirements to the 
appropriate licensee-controlled document as identified in the 
licensee's application dated December 1, 1999, and reviewed in the 
staff's safety evaluation dated August 31, 2000.
    Amendment No.: 33.
    Facility Operating License No. DPR-7: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 12, 2000 (65 FR 
1927).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 31, 2000.
    No significant hazards consideration comments received: No.

PECO Energy Company, Public Service Electric and Gas Company Delmarva 
Power and Light Company, and Atlantic City Electric Company, Docket 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: July 1, 1999, as supplemented 
August 11 and September 1, 1999.
    Brief description of amendments: The amendments revise the licenses 
to reflect changes related to the transfer of the license for the Peach 
Bottom Atomic Power Station, Units 2 and 3, to the extent held by 
Public Service Electric and Gas Company, to PSEG Nuclear Limited 
Liability Company.
    Date of issuance: August 21, 2000.
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendments Nos.: 234 and 238.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the License.
    Date of initial notice in Federal Register: August 5, 1999 (64 FR 
42728).

[[Page 56963]]

The August 11 and September 1, 1999, supplements 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 February 16, 2000.
    No significant hazards consideration comments received: No.

PECO Energy Company, PSEG Nuclear LLC, Delmarva Power and Light 
Company, and Atlantic City Electric Company; Docket Nos. 50-277 and 50-
278, Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, York County, 
Pennsylvania

    Date of application for amendments: May 31, 2000, as supplemented 
August 18, 2000.
    Brief description of amendments: The amendments revise the Peach 
Bottom Atomic Power Station, Units 2 and 3, Technical Specifications 
Surveillance Requirement 3.6.1.3.11 to allow a representative sample of 
reactor instrumentation line excess flow check valves (EFCVs) to be 
tested every 24 months, instead of testing each EFCV every 24 months.
    Date of issuance: September 8, 2000.
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendments Nos.: 235 & 239.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 9, 2000 (65 FR 
48756). The August 18, 2000, letter provided 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 September 8, 2000.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of application for amendment: April 6, 2000.
    Brief description of amendment: This amendment eliminates the 
response time testing of the Reactor Trip System and the Engineered 
Safety Feature Actuation System.
    Date of issuance: August 29, 2000.
    Effective date: August 29, 2000.
    Amendment No.: 146.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 3, 2000 (65 FR 
25768).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 29, 2000.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of application for amendment: April 6, 2000.
    Brief description of amendment: This amendment proposes to modify 
the pressure testing requirements for the American Society of 
Mechanical Engineers (ASME) Code portions of the diesel fuel oil system 
that currently require a hydrostatic test every 10 years at 110% of 
system design pressure. The revision would allow ASME Code Class 3 
portions of the diesel fuel oil system to be pressure tested in 
accordance with Section XI of the Code as required by Technical 
Specification 4.0.5. This will permit the use of Code Case N-498-1 as 
accepted by Regulatory Guide 1.147, Revision 12, for assessment of the 
diesel fuel oil system pressure boundary integrity.
    Date of issuance: August 29, 2000.
    Effective date: August 29, 2000.
    Amendment No.: 147.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 3, 2000 (65 FR 
25768).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 29, 2000.
    No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service 
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit 
No. 1, Fairfield County, South Carolina

    Date of application for amendment: January 5, 2000, as supplemented 
August 25, 2000.
    Brief description of amendment: This amendment changes Technical 
Specification (TS) 3/4 6.1.6, including its Bases, and adds TS 6.8.4.h. 
The changes support the new requirements of 10 CFR 50.55a, which 
require licensees to update their Containment Vessel Structural 
Integrity Programs to incorporate the provisions of ASME Section XI, 
Subsection IWL (1992 Edition with 1992 Addenda) and the five additional 
provisions found in 10 CFR 50.55a(b)(2)(viii).
    Date of issuance: September 6, 2000.
    Effective date: September 6, 2000.
    Amendment No.: 148.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 23, 2000 (65 
FR 9010). The August 25, 2000, supplement revised the proposed wording 
of Bases Section 3/4.6.1.6 and TS 6.8.4.h to clarify the reporting 
requirements; clarification did not impact the initial no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 6, 2000.
    No significant hazards consideration comments received: No.

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 application for amendments: November 8, 1999 (PCN-454), as 
supplemented March 16 and May 24, 2000.
    Brief description of amendments: The amendments revise Surveillance 
Requirement (SR) 3.8.1.18 of Technical Specification (TS) 3.8.1, ``A.C. 
Sources--Operating.'' The amendments revise the SR to read: Verify the 
timing of each sequenced load block is within its timer setting plus or 
minus 10% or plus or minus 2.5 seconds, whichever is greater, with the 
exception of the 5 second load group which is minus 0.5, plus 2.5 
seconds, for each programmed time interval load sequence.
    Date of issuance: September 1, 2000.
    Effective date: September 1, 2000, to be implemented within 30 days 
of issuance.
    Amendment Nos.: Unit 2-169; Unit 3-160.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 1, 1999 (64 FR 
67339).
    The supplemental letters dated March 16 and May 24, 2000, provided 
clarifying information that was within the scope of the original 
application and Federal Register notice and 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

[[Page 56964]]

Safety Evaluation dated September 1, 2000.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of application for amendments: June 1, 2000.
    Brief description of amendments: The amendments revise the reactor 
vessel pressure and temperature limit curves that are in the Technical 
Specifications.
    Date of issuance: August 29, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 222 and 163.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 28, 2000 (65 FR 
39960).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 29, 2000.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: June 22, 2000.
    Brief description of amendments: These amendments revise the 
Technical Specifications (TS) to remove the applicability of core 
alteration requirements from those TS that are designed to mitigate the 
consequences of a fuel handling accident. The applicable TS bases are 
also revised.
    Date of issuance: August 28, 2000.
    Effective date: August 28, 2000.
    Amendment Nos.: 260 and 251.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TS.
    Date of initial notice in Federal Register: July 26, 2000 (65 FR 
46017).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 28, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: April 10, 2000, as supplemented 
August 9, 2000.
    Brief description of amendment: Revision of Technical. 
Specifications (TS) to allow use of the F-star (F*) alternate repair 
criterion for degraded steam generator tubes.
    Date of issuance: September 8, 2000.
    Effective date: September 8, 2000.
    Amendment No.: 27.
    Facility Operating License No. NPF-90: Amendment revises the TS.
    Date of initial notice in Federal Register: May 31, 2000 (65 FR 
34750).
    The August 9, 2000, letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 8, 2000.
    No significant hazards consideration comments received: No.

TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas.

    Date of amendment request: May 25, 2000.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) to allow certain reactor containment 
building penetrations to be open during refueling activities under 
appropriate administrative controls. Specifically, this revision fully 
adopts the NRC-approved TS Task Force (TSTF) Traveler TSTF-312, 
Revision 1, by adding a Note to TS 3.9.4.c denoting this provision, to 
clarify the use of this allowance.
    Date of issuance: September 5, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 78 and 78.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the TSs.
    Date of initial notice in Federal Register: July 12, 2000 (65 FR 
43053).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 5, 2000.
    No significant hazards consideration comments received: No.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont 
Yankee Nuclear Power Station, Vernon, Vermont.

    Date of application for amendment: December 14, 1999.
    Brief description of amendment: The amendment relocates procedural 
details related to the Radiological Environmental Technical 
Specifications (TSs) to certain licensee-controlled documents.
    Date of Issuance: August 24, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 193.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 9, 2000 (65 FR 
6412).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated August 24, 2000.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 14th day of September 2000.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-24021 Filed 9-19-00; 8:45 am]
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