[Federal Register Volume 65, Number 96 (Wednesday, May 17, 2000)]
[Notices]
[Pages 31354-31368]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-12302]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from April 22, 2000, through May 5, 2000. The 
last biweekly notice was published on May 3, 2000 (65 FR 25761).

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; (2) create the possibility of a 
new or different kind of accident from any accident previously 
evaluated; or (3) involve a significant reduction in a margin of 
safety. The basis for this proposed determination for each amendment 
request is shown below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By June 16, 2000, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC, and electronically from 
the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room). If a request for a hearing 
or petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be

[[Page 31355]]

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

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of amendments request: May 26, 1999, as supplemented March 31, 
2000.
    Description of amendments request: The licensee proposes to change 
the allowable values in Technical Specification Section 3.3.1, Table 
3.3.1-1, Item 12, ``Reactor Coolant Flow, Steam Generator No. 1-Low'' 
and Item 13, ``Reactor Coolant Flow, Steam Generator No. 2-Low,'' to 
reduce the demonstrated spurious trip hazard associated with this 
setpoint. This application was originally noticed in the Federal 
Register on June 30, 1999 (64 FR 35201).
    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:
    Standard 1--Does the proposed change involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. The proposed change will change the Reactor Protection 
System (RPS) reactor coolant flow trip setpoints. The RPS functions 
to mitigate the consequences of an accident. The changes to the low 
reactor coolant flow trip setpoints will reduce or eliminate 
unnecessary challenges to the RPS. Therefore, the proposed change 
will not involve a significant increase in the probability of an 
accident previously evaluated.
    These changes will result in an increased time delay for the RPS 
low reactor coolant flow trip. The reanalysis of the affected 
Updated Final Safety Analysis Report (UFSAR) Chapter 15 event (UFSAR 
15.3.4, Reactor Coolant Pump Shaft Break with Loss of Offsite 
Power), with the increased time delay, shows that the dose 
consequences for this event remains bounded by the UFSAR analysis. 
Therefore, this change does not involve a significant increase in 
the consequences of an accident previously evaluated.
    Standard 2--Does the proposed change create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?
    No. The proposed change will change the RPS reactor coolant flow 
trip setpoints. The RPS functions to mitigate the consequences of an 
accident. The changes to the low reactor coolant flow trip setpoints 
will reduce or eliminate unnecessary challenges to the RPS. The 
proposed change only changes the mitigating actions of the RPS, 
without changing the required function of the RPS. Therefore, the 
change to the low reactor coolant flow trip setpoints does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    Standard 3--Does the proposed change involve a significant 
reduction in a margin of safety?
    No. The proposed change will change the RPS reactor coolant flow 
trip setpoints. The reanalysis of the affected UFSAR Chapter 15 
event (UFSAR 15.3.4, Reactor Coolant Pump Shaft Break with Loss of 
Offsite Power), with the revised reactor coolant flow trip 
setpoints, shows that the minimum departure from nucleate boiling 
ratio (DNBR) and specified acceptable fuel design limits (SAFDLs) 
for this event remains bounded by the UFSAR analysis. 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 
that review, it appears that the three standards of 10 CFR 50.92(c) are

[[Page 31356]]

satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999 NRC Section Chief: Stephen 
Dembek

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant (BSEP), Units 1 and 2, Brunswick County, 
North Carolina

    Date of amendments request: April 26, 2000.
    Description of amendments request: The proposed amendments would 
increase the maximum average ultimate heat sink (UHS) temperature 
allowed by Technical Specification (TS) 3.7.2, ``Service Water System 
and Ultimate Heat Sink.''
    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 with the maximum 24 hour average UHS water 
temperature as high as 90.5 deg.F does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The BSEP SW [Service Water] system is designed to provide 
cooling water for the removal of heat from equipment required for a 
safe reactor shutdown following a Design Basis Accident (DBA) or 
transient. This equipment includes the Diesel Generators (DGs), 
Residual Heat Removal (RHR) pump seal coolers, room cooling units 
for Emergency Core Cooling System (ECCS) equipment, and Residual 
Heat Removal Service Water (RHRSW) heat exchangers. The SW system 
also provides cooling to other components, as required, during 
normal operation. The SW system is not an initiator of any 
previously evaluated accident. The safety related components 
associated with SW cooling have been analyzed for a maximum UHS 
temperature of 92 deg.F. The proposed change maintains this maximum 
UHS temperature. As such, the qualification of safety related 
components is not affected. Therefore, the probability of occurrence 
of a previously evaluated accident is not increased.
    The new maximum 24 hour average UHS water temperature limit of 
90.5 deg. F has been evaluated and it was determined that the SW 
system will maintain sufficient heat removal capability. Existing TS 
operability requirements for the UHS ensure that conservatively 
bounding assumptions used in the analysis of the SW system's heat 
removal capability will be met, or the UHS will be declared 
inoperable. As such, the consequences of previously analyzed 
accidents are not affected[.]
    2. Operation with the maximum 24 hour average UHS water 
temperature as high as 90.5 deg.F will not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    Increasing the maximum 24 hour average UHS water temperature 
does not create the possibility of an accident of a different type 
than any evaluated previously in the safety analysis report. UHS 
water temperature does not represent an accident initiator. There is 
no physical change to any plant structure, system, or components. 
Therefore, there is no possibility of an accident of a different 
type.
    Increasing the maximum 24 hour average UHS water temperature 
does not create the possibility of a malfunction of a different type 
than any evaluated previously. The safety related components 
associated with SW cooling have been analyzed for a maximum UHS 
temperature of 92 deg.F. This maximum UHS temperature is maintained 
by the proposed change. As such, this condition does not introduce 
the possibility of a malfunction of a different type than any 
evaluated.
    3. Operation with the maximum 24 hour average UHS water 
temperature as high as 90.5 deg.F does not involve a significant 
reduction in a margin of safety.
    UHS temperature limits are established to ensure that the SW 
system is able to provide sufficient cooling water for the removal 
of heat from equipment, such as the DGs, RHR pump seal coolers, ECCS 
room cooling units, and RHRSW heat exchangers, required for a safe 
reactor shutdown following a DBA or transient. CP&L has performed an 
analysis which demonstrates that this capability is not reduced with 
the increased maximum 24 hour average UHS water temperature limit. 
Existing TS operability requirements for the UHS ensure that 
conservatively bounding assumptions used in the analysis of the SW 
system's heat removal capability will be met, or the UHS will be 
declared inoperable. As such, the ability of the SW system to 
perform its intended safety function is not affected and the margin 
of safety is not reduced.

    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.
    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: April 13, 2000
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications to accommodate the use of Framatome 
Cogema Fuels Mark-B11 fuel with M5 cladding.
    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 
consequence of an accident previously evaluated?
    No. The proposed change to the technical specifications and 
bases incorporate the use of Mark-B11 fuel assemblies with M5 
cladding. The analyzed events are initiated by the failure of 
specific plant structures, systems, or components. The change in 
fuel assembly design or cladding material does not impact the 
condition or performance of those structures, system, or components. 
Therefore, the proposed changes will not increase the probability of 
an accident previously evaluated.
    The accident analyses have been evaluated to address the changes 
in the fuel design and cladding material. The results of this 
evaluation demonstrate that the applicable acceptance criteria are 
met. Thus, the proposed changes will not increase the consequences 
of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    No. The proposed changes to the technical specifications are to 
support implementation of Mark-B11 fuel assemblies with M5 cladding. 
The changes in fuel design and cladding material do not alter the 
operating characteristics of the plant. In addition, the fuel 
handling equipment is compatible with the Mark-B11 fuel assembly 
design. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Involve a significant reduction in a margin of safety?
    No. The margin of safety is established through the design of 
the plant systems, structures, components, and the parameters within 
which the plant is operated. The proposed change does not involve 
any significant physical change to the plant. The primary design 
changes, which enhance nuclear, thermal-hydraulic and mechanical 
performance, include the following:
    1. Reduced diameter fuel rod,
    2. Flow mixing vanes on five of the six intermediate spacer 
grids,
    3. Improved grid restraint system, and
    4. M5 fuel rod cladding.
    The changes in fuel design and cladding material have been 
evaluated which demonstrates that all of the applicable acceptance 
criteria are met. Based on this, the proposed changes do not involve 
a significant reduction in a margin of safety.
    Duke has concluded based on the above that there are no 
significant hazards considerations involved in this request.

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

[[Page 31357]]

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.

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

    Date of amendment request: December 1, 1999.
    Description of amendment request: The proposed amendment would 
revise the standard by which GPU Nuclear tests charcoal used in 
engineered safeguards features (ESF) systems to American Society for 
Testing and Materials D3803-1989. These proposed changes are made in 
accordance with Generic Letter (GL) 99-02.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change is in accordance with NRC guidance in GL 99-
02 which states that new testing protocol is more accurate and 
demanding than older tests. The acceptance criteria for charcoal 
efficiency has been made more stringent and there is no change to an 
operating parameter of any system, component or structure. 
Therefore, the probability of occurrence of the consequences of an 
accident previously evaluated in the SAR [Safety Analysis Report] 
will not increase as a result of this change.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change revises the testing standard for activated 
charcoal efficiency to a more conservative methodology while 
increasing the acceptance criteria through the application of a 
safety factor. There is no change to an operating parameter of any 
system, component, or structure. Therefore, the proposed activity 
does not create the possibility for an accident or malfunction of a 
different type than any previously identified in the SAR.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.

    The proposed change does not involve a reduction in the margin 
of safety. The change is primarily administrative, adheres to NRC 
guidance, and is more conservative than the previously employed 
standard. The change does not modify an operating parameter of any 
system, or component structure. Therefore, there is no 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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Acting Section Chief: M. Gamberoni.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of amendment request: January 13, 2000.
    Description of amendment request: The amendment would add a license 
condition that requires Maine Yankee Atomic Power Company (MYAPC) to 
implement and maintain in effect all provisions of the License 
Termination Plan (LTP). MYAPC submitted the LTP in accordance with 10 
CFR 50.82(a)(9) to demonstrate that the remainder of decommissioning 
activities will be performed in accordance with Title 10 of the Code of 
Federal Regulations, will not be inimical to the common defense or 
security or to the health and safety of the public, and will not have a 
significant effect on the quality of the environment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The requested license amendment does not authorize any plant 
activities beyond that allowed by 10 CFR Chapter I or beyond that 
considered in the DSAR [Defueled Safety Analysis Report]. The 
bounding accident described in the DSAR for potential airborne 
activity is the postulated resin cask drop accident in the Low Level 
Radioactive Waste Storage Building. This accident is expected to 
contain more potential airborne activity than can be released from 
other decommissioning events. The radionuclide distribution assumed 
for the spent resin cask has more transuranics (the major dose 
contributor) than the distribution in the components involved in 
other decommissioning accidents. The accidents considered in the 
DSAR include: 1) Explosion of Liquid Petroleum Gas (LPG) Leaked from 
a Front End Loader or Forklift, 2) Explosion of Oxyacetylene During 
Segmenting of the Reactor Vessel Shelf, 3) Release of Radioactivity 
from the RCS Decontamination Ion Exchange Resins, 4) Gross Leak 
During In-Situ Decontamination, 5) Segmentation of RCS Piping with 
Unremoved Contamination, 6) Fire Involving Contaminated Clothing or 
Combustible Waste, 7) Loss of Local Airborne Contamination Control 
During Blasting or Jackhammer Operations, 8) Temporary Loss of 
Services, 9) Dropping of Contaminated Concrete Rubble, 10) Natural 
Phenomena and 11) Transportation Accidents. The probabilities and 
consequences for these accidents are estimated in the basis 
documentation for DSAR Section 7. No systems, structures, or 
components that could initiate or be required to mitigate the 
consequences of an accident are affected by the proposed change in 
any way not previously evaluated in the DSAR. Since Maine Yankee 
does not exceed the salient parameters associated with the plant 
referenced in the basis documentation in any material respects, it 
is concluded that these probabilities and consequences are not 
increased. Therefore, the proposed change to the Maine Yankee 
License does not involve any increase in the probability or 
consequences of any accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The requested license amendment does not authorize any plant 
activities which could precipitate or result in any accidents beyond 
that considered in the DSAR. The accidents previously evaluated in 
the DSAR are described above. These accidents are described in the 
basis documentation for DSAR Section 7. The proposed change does not 
affect plant systems, structures, or components in any way not 
previously evaluated in the DSAR. Since Maine Yankee does not exceed 
the salient parameters associated with the plant referenced in the 
basis documentation in any material respects, it is concluded that 
these accidents appropriately bound the kinds of accidents possible 
during decommissioning. Therefore, the proposed change to the Maine 
Yankee License would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety defined in Maine Yankee's license basis for 
the consequences of decommissioning accidents has been established 
as the margin between the bounding decommissioning accident and the 
dose limits associated with the need for emergency plan offsite 
protection, namely the Environmental Protection Agency Protective 
Action Guidelines EPA-PAGs. As described above, the bounding 
decommissioning accident is the postulated resin cask drop accident 
in the Low Level

[[Page 31358]]

Radioactive Waste Storage Building. Since the bounding 
decommissioning accident is expected to contain more potential 
airborne activity than can be released from other decommissioning 
events and since the radionuclide distribution assumed for the spent 
resin cask has more transuranics (the major dose contributor) than 
the distribution in the components involved in other decommissioning 
accidents, the margin of safety associated with the consequences of 
decommissioning accidents cannot be reduced. The margin of safety 
defined in the statements of consideration for the final rule on the 
Radiological Criteria for License Termination is described as the 
margin between the 100 mrem/yr public dose limit established in 10 
CFR 20.1301 for licensed operation and the 25 mrem/yr dose limit to 
the average member of the critical group at a site considered 
acceptable for unrestricted use. This margin of safety accounts for 
the potential effect of multiple sources of radiation exposure to 
the critical group. Since the license termination plan was designed 
to comply with the radiological criteria for license termination for 
unrestricted use, the margin of safety cannot be reduced. Therefore, 
the proposed changes to the Maine Yankee License would not involve a 
significant reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendment involves no significant hazards consideration.
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, 321 Old Ferry Road, Wiscasset, Maine 04578.
    NRC Section Chief: Michael T. Masnik.

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

    Date of amendment request: November 30, 1999, as supplemented on 
April 28, 2000.
    Description of amendment request: The licensee proposes to change 
the technical specifications (TSs) relating to the emergency diesel 
generator fuel sampling/testing surveillance requirements (SRs). The 
changes would provide a new administrative control to establish, 
implement, and maintain a diesel fuel oil testing program, relocate 
fuel oil sampling/testing surveillance requirements and fuel oil 
storage tank cleaning frequency requirement to a new diesel fuel oil 
testing program which will reside in the Seabrook Station Technical 
Requirements (SSTR) Manual. The change will also add references to the 
A.C. Sources--Shutdown surveillance requirement to perform additional 
activities while in modes 5 and 6.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components (SSCs) to 
perform their intended function to mitigate the consequences of an 
initiating event within the acceptance limits assumed in the Updated 
Final Safety Analysis Report (UFSAR).
    The proposed changes do not affect the source term, containment 
isolation or radiological release assumptions used in evaluating the 
radiological consequences of an accident previously evaluated in the 
Seabrook Station UFSAR. Further, the proposed changes do not 
increase the types and amounts of radioactive effluent that may be 
released offsite, nor significantly increase individual or 
cumulative occupational/public radiation exposures. The proposed 
change to SR 4.8.1.2 provides additional requirements for operation 
of the facility. These additional requirements are not initiators of 
analyzed events and will not alter assumptions relative to 
mitigation of accident or transient events. The proposed change does 
not adversely affect previously evaluated accidents.
    Therefore, the proposed changes do 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 accident previously evaluated.
    The proposed changes do not involve physical alteration of plant 
SSCs or changes in parameters governing the manner in which the 
plant is operated and maintained in a state of readiness. The 
changes do not introduce a new mode of plant operation.
    As discussed in the above narrative, the proposed change to SR 
4.8.1.2 provides additional requirements for operation of the 
facility. These additional requirements are not initiators of 
analyzed events and will not alter assumptions relative to 
mitigation of accident or transient events. The proposed change does 
not adversely affect previously evaluated accidents.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes do not involve a reduction in a margin of 
safety because they do not adversely affect assumptions used in 
transient or safety analyses. The details associated with the 
involved specifications are not required to be in the TS to provide 
adequate protection of the public health and safety, since the TS 
still retains the requirement for compliance with the applicable 
standards. The level of safety of facility operation is unaffected 
by the changes since there is no change in the intent of the TS 
requirements of ensuring fuel oil is of the appropriate quality for 
diesel generator use.
    The proposed change to the A.C. Sources--Shutdown SR imposes an 
additional level of requirements that are more restrictive than the 
current TS requirements for operation of the facility in Modes 5 and 
6. The additional requirements being proposed enhance assurance that 
the same fuel oil quality requirements are met, and visual 
inspection activities conducted, whenever a diesel generator is 
required to be OPERABLE.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    Based on this review, it appears that the three standards of 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: April 14, 2000.
    Description of amendment request: The licensee proposes to relocate 
Technical Specification (TS) Sections TS 3/4.9.5, ``Communications'', 
TS 3/4.9.6, ``Refueling Machine'', and TS 3/4.9.6, ``Crane Travel--
Spent Fuel Storage Area'' to the Seabrook Station Technical 
Requirements Manual.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes to relocate Technical Specifications 3/
4.9.5, 3/4.9.6 and 3/4.9.7 to the Technical Requirements Manual 
(TRM) are administrative in nature and do not adversely affect 
accident initiators or precursors nor alter the design assumptions, 
conditions, configuration of the

[[Page 31359]]

facility or the manner in which it is operated. The proposed changes 
do not alter or prevent the ability o[f] structures, systems, or 
components to perform their intended function to mitigate the 
consequences of an initiating event within the acceptance limits 
assumed in the Updated Final Safety Analysis Report [UFSAR].
    The subject specifications relocated to the Technical 
Requirements Manual will continue to be administratively controlled. 
The TRM is a licensee-controlled document, which contains certain 
technical requirements and is the implementing manual for the 
Technical Specification Improvement Program. Changes to these 
requirements are reviewed and approved in accordance with Seabrook 
Station Technical Specification, Section 6.7.1.i, and as outlined in 
the TRM.
    Therefore, the proposed change 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 accident previously evaluated.
    The proposed change does not alter the design assumptions, 
conditions, or configuration of the facility or the manner in which 
the plant is operated. There are no changes to the source term or 
radiological release assumptions used in evaluating the radiological 
consequences in the Seabrook Station UFSAR. The proposed change has 
no adverse impact on component or system interactions. The proposed 
change will not adversely degrade the ability of systems, structures 
and components important to safety to perform their safety function 
nor change the response of any system, structure or component 
important to safety as described in the Seabrook Station Updated 
Final Safety Analysis Report (UFSAR). The proposed changes are 
administrative in nature and do not change the level of programmatic 
and procedural details of assuring operation of the facility in a 
safe manner. Since there are no changes to the design assumptions, 
conditions, configuration of the facility, or the manner in which 
the plant is operated and surveilled, the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    There is no adverse impact on equipment design or operation and 
there are no changes being made to the Technical Specification 
required safety limits or safety system settings that would 
adversely affect plant safety. The proposed change is administrative 
in nature and does not reduce the level of programmatic or 
procedural controls associated with the activities presently 
performed via Technical Specifications 3/4.9.5, 3/4.9.6 and 3/4.9.7.
    Future changes to the subject technical requirements will be 
reviewed and approved in accordance with Seabrook Station Technical 
Specification, Section 6.7, and as outlined in the Technical 
Requirements Manual. Specifically, all changes to the Technical 
Requirements Manual require a 10 CFR 50.59 safety evaluation and 
will be reviewed and approved by the Station Operations Review 
Committee (SORC) prior to implementation.
    Therefore, relocation of the requirements contained in Technical 
Specifications 3/4.9.5, 3/4.9.6 and 3/4.9.7 to the Technical 
Requirements Manual does not involve a significant reduction in the 
margin of safety provided in the existing specifications.

    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: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: November 24, 1999, as supplemented by 
letter dated February 10, 2000.
    Description of amendment request: The amendment will establish 
charcoal filter testing requirements in the technical specifications 
(TSs) for the Auxiliary Building Ventilation (ABV) System, the Control 
Room Envelope Air Conditioning System (CREACS), and the Fuel Handling 
Building Ventilation (FHV) System that are consistent with Generic 
Letter 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal,'' 
dated June 3, 1999.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    The operation of the Salem units in accordance with the proposed 
changes will not involve a significant increase in the probability or 
consequences of an accident previously evaluated. The proposed TS 
changes do not involve any physical changes to plant structures, 
systems, or components (SSC). The FHV, CREACS, and ABV systems will 
continue to function as designed. The FHV, CREACS, and ABV systems are 
designed to mitigate the consequences of an accident. The proposed 
changes also will not affect the sequence of any accidents previously 
analyzed. The proposed TS surveillance requirement changes implement 
testing methods that demonstrate charcoal filter capability and 
establish acceptance criteria, which ensure that Salem's design basis 
assumptions continue to be met. The proposed surveillance requirement 
acceptance criteria ensure that the FHV, CREACS, and ABV safety 
functions will be accomplished. Therefore, the proposed TS changes 
would not result in a significant increase of the consequences of an 
accident previously evaluated, nor do they involve an increase in the 
probability of an accident previously evaluated.
    The operation of the Salem units in accordance with the proposed 
changes does not create the possibility of a new or different kind of 
accident from any accident previously evaluated. The proposed TS 
changes do not involve any physical changes to the design of any plant 
SSC. The design and operation of the FHV, CREACS, and ABV systems are 
not changed from those currently described in Salem's licensing basis. 
The FHV, CREACS, and ABV systems will continue to function as designed 
to mitigate the consequences of an accident. Implementing the proposed 
charcoal filter testing methods and acceptance criteria does not change 
the operation of the FHV, CREACS, and ABV systems that would create a 
different type of accident previously evaluated. In addition, the 
proposed TS changes do not alter the conclusions described in Salem's 
licensing basis regarding the safety-related functions of these 
systems. Therefore, the proposed TS changes do not create the 
possibility of a new or different kind of accident from any previously 
evaluated.
    The operation of the Salem units in accordance with the proposed 
license amendment will not be changed nor result in a significant 
reduction to margins of safety. The licensee is not proposing any 
modifications to FHV, CREACS, and ABV systems design or operation, and 
there are no changes being made to the TS-required safety limits or 
safety system settings that would adversely affect plant safety. The 
proposed changes modify the TSs to reference appropriate test 
parameters for performing laboratory testing of nuclear-grade charcoal 
in engineered safety feature filtration systems in accordance with ASTM 
D3803-1989. The imposition of the more conservative charcoal filter 
testing requirements associated with ASTM D3803-1989 will not involve a 
significant reduction in a margin of safety.
    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.

[[Page 31360]]

    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment requests: April 20, 2000 (PCN-503).
    Description of amendment requests: The amendment application 
proposes to revise the San Onofre Nuclear Generating Station, Units 2 
and 3, Technical Specification (TS) 5.5.2.5, ``Reactor Coolant Pump 
Flywheel Inspection Program.'' The proposed change would revise the 
required volumetric examination frequency of the upper flywheel on each 
of the primary reactor coolant pump motors from a 3-year to a 10-year 
cycle.
    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?
    Response: No.
    Missile generation from a Reactor Coolant Pump (RCP) flywheel 
could damage the reactor coolant system, the containment, or other 
equipment or systems important to safety. The fracture mechanics 
analysis performed to support the change shows that a preexisting 
flaw of an initial size at the detection threshold level will not 
grow to a flaw size necessary to create flywheel missiles within the 
life of the plant. The fracture mechanics analysis conservatively 
assumes minimum material toughness properties, maximum flywheel 
speed, location of flaw in the highest stress region of the 
flywheel, and a number of start/stop cycles eight times greater than 
the design basis. Therefore, an existing flaw in the flywheel will 
not grow to a size that exceeds the allowable flaw size for either 
normal operating or accident conditions over the plant life. On this 
basis, the extension of the 3-year interval inspection to a 10-year 
interval will not involve a significant increase in the probability 
of an accident previously considered. The proposed changes do not 
increase the amount of radioactive material available for release or 
modify any systems used for preventing or mitigating such releases 
during accident conditions. Therefore, these changes do not involve 
a significant increase in the consequences of any accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated?
    Response: No.
    The proposed changes will not change the design configuration, 
or method of operation of the plant. Therefore, the proposed changes 
will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety?
    Response: No.
    Significant conservatisms have been used in the calculation of 
allowable flaw size (critical flaw size) and flaw growth for each 
RCP flywheel design. These include minimum fracture toughness 
properties, code reference crack growth rate curves, maximum 
flywheel accident speed, postulated flaw location at the highest 
stress region of the flywheel, and a number of start/stop cycles 
that is eight times the number expected in a plant life. The final 
flaw size has been determined to remain smaller than the allowable 
flaw size for the flywheel under the relevant design conditions, 
including postulated accident conditions. Therefore, the extension 
of the 3-year interval inspection to a 10-year interval will not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

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

    Date of application for amendments: February 4, 2000 (TS 99-14).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah Nuclear Plant (SQN) Technical Specification 
Limiting Conditions for Operations for the reactor coolant system cold 
leg accumulators (CLAs). The upper CLA water limit and required 
pressure range would both be decreased to more appropriately account 
for instrument uncertainties and instrument line tap locations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), 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.
    The capability of the accumulators to perform their safety 
function is not affected by this change. All components and system 
functional requirements remain the same. There are no new sequences 
of events which would increase the probability of an accident 
analyzed in the Final Safety Analysis Report (FSAR). Therefore, the 
proposed activity does not increase the probability of an accident 
previously evaluated in the FSAR. The fuel cladding peak temperature 
established by the ECCS [Emergency Core Cooling System] evaluation 
model remains below 2200 degrees Fahrenheit for a loss-of-coolant 
accident (LOCA). As such, the assumptions on fuel failure and 
isotope release post-LOCA do not change from the information 
presented in the FSAR.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The primary function of the CLAs is in the event of a large 
break LOCA to support accident mitigation. CLAs are not a 
contributor to events that could generate accidents. The CLA system 
volume capability bounds this change in operational limits and the 
system is not physically changing. Therefore, the proposed activity 
does not create a possibility for an accident of a different type 
than any evaluated previously.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The safety function provided by the CLAs is to inject core 
cooling water into the reactor coolant system when system pressure 
decreases below a predetermined value during a LOCA. The timing 
(function of pressure) and amount (function of volume) of cooling 
water is modeled in the ECCS evaluation model. The proposed changes 
to the accumulator operational limits have been evaluated using the 
Sequoyah plant specific ECCS model. The evaluation shows an increase 
in the peak fuel cladding temperature from 2162 degrees Fahrenheit 
to 2185 degrees Fahrenheit. The results confirm that existing LOCA 
safety analysis acceptance criteria (established by 10 CFR 50.46) 
continue to be met for the revised accumulator limits. The safety 
analysis acceptance criteria continues to be met with the revised 
limits. The 23 degree increase in the peak fuel cladding temperature 
associated with accumulator operation is not a significant reduction 
in the margin of safety.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

[[Page 31361]]

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

    Date of amendment request: March 6, 2000 (TS 99-09).
    Description of amendment request: The proposed amendment would 
revise the Watts Bar Nuclear Plant (WBN) Unit 1 Technical 
Specifications (TS) and associated TS Bases for Limiting Condition for 
Operation (LCO) 3.9.4 Containment Penetrations. The revision would 
permit both doors of the containment personnel airlocks to be open 
during refueling operations to facilitate personnel and equipment 
access to containment. It would also allow containment penetration flow 
paths to be open under administrative controls to facilitate 
maintenance activities during refueling operations.
    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.
    The proposed change to WBN Technical Specification LCO 3.9.4, 
Refueling Operations--Containment Penetrations, would allow both 
doors of the containment personnel airlocks and certain containment 
penetration flow paths to be open during core alterations and 
movement of irradiated fuel within containment under specific 
administrative controls. The proposed change is consistent with NRC 
approved TS travelers TSTF-68, R2 and TSTF-312, R1, and proposes 
controls similar to the administrative controls currently allowed by 
WBN TS (LCO 3.6.3) for containment penetrations during more 
restrictive, higher operational modes. The administrative controls 
will ensure appropriate personnel are aware of the open personnel 
airlocks and penetration flow paths and ensure designated 
individual(s) are assigned to promptly close the airlock doors and 
penetration flow paths in the event of a fuel handling accident 
(FHA) inside containment. Timely closure of penetration flow paths 
and closure of the airlock doors following containment evacuation 
will ensure that the unlikely transmission of radioactive material 
from the reactor building to the auxiliary building is minimized.
    In order to minimize the consequences of any leakage of 
radionuclides past these open penetrations during the period of time 
before their closure, additional procedural controls will be 
provided to ensure the integrity of the WBN auxiliary building 
secondary containment enclosure (ABSCE) boundary and proper 
auxiliary building gas treatment system (ABGTS) operation. These 
controls will ensure that in the event of a fuel handling accident 
(FHA) inside containment, the following will be promptly 
accomplished: shutdown and isolation of the reactor building purge 
air ventilation system, auxiliary building isolation, and initiation 
of ABGTS. Therefore, through the use of these controls for the 
proposed license amendment, the offsite dose consequences of a FHA 
inside containment with open airlock doors and/or open penetration 
flow paths remain well within the 10 CFR 100 limits and within the 
limits of 10 CFR 50, Appendix A, General Design Criteria 19 for 
control room operator dose.
    [The licensee's application also states that ``The results for 
the fuel handling analysis inside containment with open airlock 
doors and/or open penetration flow paths are bounded by the current 
analysis.'']
    The containment personnel airlock doors and containment 
penetration flow paths are not initiators to any previously 
evaluated accident for WBN. In addition, the position of the airlock 
doors and penetration flow paths during refueling operations has no 
affect on the probability of the occurrence of any accident 
previously evaluated. The proposed revision does not alter any plant 
equipment or operating practices in such a manner that the 
probability of an accident is increased. Since the probability of a 
accident is not affected by the positions of the containment 
personnel airlock doors, and because the doses remain within 
acceptable limits, 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.
    The open containment personnel airlock doors and containment 
penetration flow paths are not accident initiators and do not 
represent a significant change in the configuration of the plant. 
The proposed allowance to open the containment personnel airlock 
doors and penetrations during refueling operations will not 
adversely affect plant safety functions or equipment operating 
practices such that a new or different accident could be created. 
Therefore, since plant safety functions are not adversely affected 
and the isolation status of containment personnel airlock doors and 
penetration flow paths do not contribute to the initiation of 
postulated accidents, the proposed revision will not create 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.
    WBN Technical Specification LCO 3.9.4 closure requirements for 
containment penetrations ensure that the consequences of a 
postulated FHA inside containment during core alterations or fuel 
handling activities remain within acceptable limits. The LCO 
establishes containment closure requirements, which limit the 
potential escape paths for fission products by ensuring that there 
is at least one integral barrier to the release of radioactive 
material. The proposed change to allow the containment personnel 
airlock doors and containment penetration flow paths to be open 
during refueling operations under administrative controls does not 
significantly affect the expected dose consequences of a FHA because 
of the absence of containment pressurization during refueling. 
Without this motive force, the potential for additional offsite dose 
consequence is unlikely. The proposed administrative controls 
provide assurance that prompt closure of the airlock doors and 
penetration flow paths will be accomplished in the event of a FHA 
inside containment thus minimizing the transmission of radioactive 
material from the reactor building to the auxiliary building. Under 
the proposed TS change, the provisions to ensure shutdown and 
isolation of the reactor building purge air ventilation system, 
auxiliary building isolation, and initiation of ABGTS and to 
promptly isolate open penetration flow paths and close the airlock 
doors following containment evacuation, provide assurance that the 
offsite dose consequences of a FHA inside containment will remain 
well within the 10 CFR 100 limits and within the limits of 10 CFR 
50, Appendix A, General Design Criteria 19 for control room operator 
dose. Therefore, the proposed change to the WBN Technical 
Specifications does not involve a significant reduction in the 
margin of safety.

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

[[Page 31362]]

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

    Date of application for amendment: September 16, 1999.
    Brief description of amendment: The proposed amendment would 
increase the licensed capacity for spent fuel assembly storage in the 
Spent Fuel Pool and revise the configuration for storage of fresh fuel.
    Date of publication of individual notice in the Federal Register: 
December 8, 1999 (64 FR 68702).
    Expiration date of individual notice: January 7, 2000.

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 amendment request: February 29, 2000, as supplemented on 
March 31, 2000.
    Brief description of amendment request: The amendments would add a 
note to the completion time of Condition A for Technical Specification 
3.7.2, ``Emergency Service Water (ESW) System and Normal Heat Sink.'' 
This note would provide a one-time extension to the completion time for 
one ESW subsystem inoperable from 7 to 14 days. This note would allow 
the replacement of one ESW pump currently scheduled to occur in May 
2000 and will expire on May 31, 2000.
    Date of publication of individual notice in Federal Register: March 
9, 2000 (65 FR 12589).
    Expiration date of individual notice: April 10, 2000.

PP&L, Inc., Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric 
Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: March 14, 2000, as supplemented March 
27, 2000. 
    Brief description of amendment request: The proposed amendment 
would amend the licenses to change the required implementation date for 
previously issued license Amendment No. 184 to Facility Operating 
License NPF-14 and Amendment No. 158 to Facility Operating License NPF-
22. The proposed amendment would not alter any of the requirements of 
the SSES Unit 1 and 2 Technical Specifications (TSs).
    Date of publication of individual notice in Federal Register: April 
27, 2000 (65 FR 24718).
    Expiration date of individual notice: May 30, 2000.

PP&L, Inc., Docket No. 50-388, Susquehanna Steam Electric Station, Unit 
2, Luzerne County, Pennsylvania

    Date of amendment request: April 10, 2000.
    Brief description of amendment request: Permits deferral of testing 
of primary containment penetration flange o-rings on spectacle flanges 
2S299A and 2S299B until the Unit 2 10th refueling outage, scheduled for 
spring 2001 or a prior Unit 2 outage requiring entry into Mode 4.
    Date of publication of individual notice in Federal Register: April 
21, 2000 (65 FR 21487).
    Expiration date of individual notice: May 22, 2000.

Notice of Issuance of Amendments to Facility Operating Licenses

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

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of application for amendment: October 25, 1999 (U-603281).
    Brief description of amendment: The amendment revised the Technical 
Specification definitions for channel calibrations, channel functional 
tests, and logic system functional tests.
    Date of issuance: April 25, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 128.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 12, 2000 (65 FR 
1920).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 25, 2000.
    No significant hazards consideration comments received: No.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of application for amendment: November 19, 1999, as 
supplemented on March 16, 2000.
    Brief description of amendment: This amendment revises the 
Technical Specifications (TS) to incorporate the American Society for 
Testing and Materials (ASTM) D3803-1989, Standard Test Method for 
Nuclear-Grade Activated Carbon,'' in accordance with NRC Generic Letter 
(GL) 99-02, ``Laboratory Testing Of Nuclear-Grade Activated Charcoal,'' 
dated June 3, 1999. Specifically, TS 4.7.6 has been revised for the 
Control Room Emergency Filtration System, TS 4.7.7 has been revised for 
the Reactor Auxiliary Building Emergency Exhaust System, and TS 4.9.12 
has been revised for the Fuel Handling Building Emergency Exhaust 
System.
    Date of issuance: May 2, 2000.
    Effective date: May 2, 2000.
    Amendment No. 98.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70081).
    The March 16, 2000, submittal contained clarifying information 
only,

[[Page 31363]]

and did not change the initial no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 2, 2000. No significant 
hazards consideration comments received: Yes. One comment was received, 
and is addressed in the above-referenced Safety Evaluation.

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

    Date of application for amendment: March 17, 2000.
    Brief description of amendment: The amendment revises Technical 
Specifications (TSs) associated with probes used in steam generator 
tube inspections, specifically TS Section 4.13.A.3.f. The proposed 
change would provide more flexibility in the type of probe used and 
would reflect current technological advances in inspection equipment, 
while still maintaining the current 610-mil diameter probe restriction. 
]
    Date of issuance: April 28, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 209.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 27, 2000 (65 FR 
16230).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 28, 2000.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Units 1 and 2, Pope County, Arkansas

    Date of amendment request: November 16, 1999.
    Brief description of amendments: The proposed changes to the 
Arkansas Nuclear One, Units 1 and 2, Technical Specifications (TSs) and 
associated Bases provided a 30-day allowed outage time (AOT) for 
startup transformer No. 2, which is an offsite power source shared by 
both units. This 30-day AOT will be used infrequently for the purpose 
of performing preventative maintenance to increase the reliability of 
the transformer. In addition, changes have been made to the 
requirements associated with demonstrating the operability of the 
emergency diesel generators (EDGs), in the event a required power 
source is inoperable, to increase the reliability of the EDGs.
    Date of issuance: April 28, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 206 and 215.
    Facility Operating License Nos. DPR-51 and NPF-6: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4271).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 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: February 24, 2000.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 4.4.11 on reactor coolant system vent flow 
verification, TS 4.6.1.1.a on containment penetration closure 
verification (non-automatic), and TS 4.6.3.1.2 on containment isolation 
valve actuation verification. The changes eliminated unnecessary mode 
restrictions on these surveillance requirements.
    Date of issuance: April 26, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 214.
    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 April 26, 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 6, 1998, as supplemented by 
letter dated March 3, 2000, Moderator Temperature Coefficient test near 
the end of each cycle.
    Brief description of amendment: The proposed change modifies the 
requirement to perform a Moderator Temperature Coefficient test near 
the end of each cycle.
    Date of issuance: April 21, 2000.
    Effective date: As of the date of issuance and shall be implemented 
60 days from the date of issuance.
    Amendment No.: 159.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46435). The March 3, 2000, letter did not change the scope of the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 21, 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 January 6, 2000.
    Brief description of amendment: The proposed change modifies plant 
technical specifications to extend the Reactor Coolant System Pressure 
Temperature Curve Limit to 16 Effective Full Power Years.
    Date of issuance: April 24, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 160.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4276).
    The January 6, 2000, letter reduced EFPY from 20 years, requested 
in the July 15, 1999, letter, to 16 years. This change is bounded by, 
and did not change the scope of, the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 24, 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: November 23, 1999, as 
supplemented February 22, 2000.
    Brief description of amendments: The amendments make the following 
changes to the Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 
and BVPS-2) Technical Specifications (TSs): (1) For BVPS-1, 
surveillance requirement (SR) 4.8.1.1.2.b.3.b is revised to reflect a 
narrower required diesel generator (DG) frequency band; an associated 
footnote is deleted; associated Bases are revised to reflect these TS 
changes. (2) For BVPS-2, SR 4.8.1.1.2.f is revised to clarify that the

[[Page 31364]]

DGs are only required to achieve a minimum frequency and voltage within 
the first 10 seconds of the related test, and that the stated voltage 
and frequency bands are requirements for steady state operation of the 
DGs; a footnote is also added to this SR. (3) Page formats are revised 
as needed to permit the addition or deletion of text.
    Date of issuance: April 25, 2000.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 230 and 109.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 8, 2000 (65 FR 
12292). The February 22, 2000, letter provided supplemental information 
and did not change the initial proposed no significant hazards 
consideration determination or expand the amendments beyond the scope 
of the initial notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 25, 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: September 22, 1999, as 
supplemented April 27, 2000.
    Brief description of amendment: The amendment allowed a one-time 
only extension to the surveillance interval of the Technical 
Specification Surveillance 4.7.12.d for functional testing of snubbers. 
The extension is limited to the first re-entry into MODE 6 following 
the defueled condition during the 8th refueling outage or November 30, 
2000, whichever occurs sooner.
    Date of issuance: May 3, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 110.
    Facility Operating License No. NPF-73: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 17, 1999, (64 
FR 62711).
    The April 27, 2000, letter 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 May 3, 2000.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: November 2, 1999.
    Brief description of amendment: This amendment revised the 
Technical Specifications (TSs) to modify 1) TS Table 3.3-4, ``Safety 
Features Actuation System Instrumentation Trip Setpoints,'' to remove 
the ``Trip Setpoint'' values for Instrument String Functional Unit 
``f'', Borated Water Storage Tank (BWST) Level, 2) the ``Allowable 
Values'' entry for this same Functional Unit, consistent with updated 
calculations using current setpoint methodology, 3) TS 3/4.3.2.1, 
``Safety Features Actuation System Instrumentation,'' and Bases to 
reflect the removal of ``Trip Setpoints'' described above, and 4) TS 3/
4.5.4, ``Emergency Core Cooling Systems--Borated Water Storage Tank,'' 
and Bases to increase the minimum volume of water in the BWST.
    Date of issuance: May 4, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 241.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70087).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 4, 2000.
    No significant hazards consideration comments received: No.

Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit 
1, DeWitt County, Illinois

    Date of application for amendment: October 23, 1998, as 
supplemented February 22 and June 24, 1999, and March 31, 2000.
    Brief description of amendment: The amendment would allow 
implementation of a feedwater leakage control system to address leakage 
through the primary containment feedwater penetration valve.
    Date of issuance: April 25, 2000.
    Effective date: April 25, 2000.
    Amendment No.: 127.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 64118).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 25, 2000.
    No significant hazards consideration comments received: No.

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 application for amendments: February 18, 2000, as 
supplemented March 31, 2000.
    Brief description of amendments: The proposed license amendments 
would approve a change to the facility involving an unreviewed safety 
question discovered by the licensee during a 10 CFR 50.59 evaluation of 
modifications to the auxiliary feedwater (AFW) pump rooms to protect 
the equipment in the rooms from the environmental effects of a 
postulated high-energy line break. This will be accomplished by sealing 
the AFW pump rooms to ensure that the rooms do not communicate with the 
turbine buildings or each other.
    Date of issuance: April 25, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 244.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Operating License.
    Date of initial notice in Federal Register: February 25, 2000 (65 
FR 10116).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 25, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 13, 1999.
    Description of amendment request: The amendment changes the license 
to delete expired license conditions and to make editorial and 
administrative changes to correct or clarify the license.
    Date of issuance: April 27, 2000.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 68.
    Facility Operating License No. NPF-86: Amendment revised the 
License.
    Date of initial notice in Federal Register: February 9, 2000 (65 FR 
6408).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 27, 2000.
    No significant hazards consideration comment received: No.

[[Page 31365]]

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

    Date of amendment request: February 18, 2000.
    Description of amendment request: The amendment revises Technical 
Specifications (TSs) Surveillance Requirements 4.0.5.a, 4.0.5.b, 
4.0.5.e, and 4.4.6.2.2.e. These changes are required to ensure 
consistency between the TSs and the second 10-year inservice test 
program by approval to use the 1995 Edition and 1996 Addenda of the 
American Society of Mechanical Engineers (ASME) Code for Operation and 
Maintenance of Nuclear Power Plants (OM Code). The revision to TSs 
Surveillance Requirement 4.0.5.a also incorporates semi-quarterly and 
biennial intervals to the list of required frequencies for performing 
inservice test and inspection activities.
    Date of issuance: May 8, 2000.
    Effective date: As of its date of issuance, and shall be 
implemented by August 18, 2000.
    Amendment No.: 69.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 5, 2000 (65 FR 
17917).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 8, 2000.
    No significant hazards consideration comments received: No.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: December 14, 1999, as 
supplemented February 11, March 30, and April 26, 2000.
    Brief description of amendment: This amendment will revise 
Technical Specifications (TSs) Sections: 3.3.2.1, ``Instrumentation--
Engineered Safety Feature Actuation System Instrumentation;'' 3.3.3.1, 
``Instrumentation--Monitoring Instrumentation--Radiation Monitoring;'' 
3.7.6.1, ``Plant Systems--Control Room Emergency Ventilation System;'' 
3.9.3.1, ``Refueling Operations--Decay Time;'' 3.9.4, ``Refueling 
Operations--Containment Penetrations;'' 3.9.9, ``Refueling Operations--
Containment Radiation Monitoring;'' 3.9.10 ``Refueling Operations--
Containment Purge Valve Isolation System;'' 3.9.13, ``Refueling 
Operations--Storage Pool Radiation Monitoring;'' 3.9.14, ``Refueling 
Operations--Storage Pool Area Ventilation System--Fuel Movement;'' 
3.9.15, ``Refueling Operations--Storage Pool Area Ventilation System--
Fuel Storage;'' 3.9.16.1, ``Refueling Operations--Shielded Cask;'' 
3.9.16.2, ``Refueling Operations--Shielded Cask;'' 3.9.17, ``Refueling 
Operations--Movement of Fuel in Spent Fuel Pool;'' and 3.9.19.2, 
``Refueling Operations--Spent Fuel Pool--Storage Pattern''; and add new 
TS 3.3.4, ``Containment Purge Valve Isolation Signal.'' The requested 
changes would make the TSs and the Final Safety Analysis Report (FSAR) 
consistent with the new analyses of the fuel handling and cask drop 
accidents. The Index Pages and the Bases for these TSs will be modified 
to reflect these proposed changes.
    Date of issuance: April 28, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 245.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 17, 2000 (65 FR 
14632).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 28, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 15, 1999, as supplemented 
July 20, September 3, and November 29, 1999, and January 18, 2000.
    Brief description of amendment: The amendment modifies the license 
to change the number of owners from 14 to 13 and to remove Montaup 
Electric Company as an owner as a result of the transfer of its 
interest in Millstone Nuclear Power Station, Unit No. 3 to New England 
Power Company, an existing owner.
    Date of issuance: May 1, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 180.
    Facility Operating License No. NPF-49: Amendment revised the 
License.
    Date of initial notice in Federal Register: January 19, 2000 (65 FR 
2990).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 24, 2000, issued with the 
February 24, 2000, Order approving the transfer as noticed in the 
Federal Register on March 1, 2000 (65 FR 11091).
    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: February 29, 2000, as 
supplemented on March 31, 2000.
    Brief description of amendments: The amendments will add a note to 
the completion time of Condition A for Technical Specification 3.7.2, 
``Emergency Service Water (ESW) System and Normal Heat Sink.'' This 
note will provide a one-time extension to the completion time for one 
ESW subsystem inoperable from 7 to 14 days. This note will allow the 
replacement of one ESW pump currently scheduled to occur in May 2000 
and will expire on May 31, 2000.
    Date of issuance: April 25, 2000.
    Effective date: Both units, as of the date of issuance and shall be 
implemented no later than May 31, 2000.
    Amendments Nos.: 231 and 236.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 9, 2000 (65 FR 
12589).
    The March 31, 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 April 25, 2000.
    No significant hazards consideration comments received: No.

PP&L, Inc., Docket No. 50-388, Susquehanna Steam Electric Station, Unit 
2, Luzerne County, Pennsylvania

    Date of application for amendment: April 10, 2000.
    Brief description of amendment: The amendment adds a note to 
Technical Specification Surveillance Requirement 3.6.1.1.1 to defer 
performance of this test on a one-time basis for spectacle flanges 
2S299A and 2S299B o-rings until the Unit 2 10th Refueling Outage 
(Spring 2001) or a prior Unit 2 outage requiring entry into Mode 4. The 
change allowed Unit 2 operation to continue until an outage occurs 
where leak rate surveillance testing on spectacle flanges 2S299A and 
2S299B can be performed.
    Date of issuance: May 8, 2000.

[[Page 31366]]

    Effective date: As of date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 160.
    Facility Operating License No. NPF-22. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 21, 2000 (65 FR 
21487).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 8, 2000.
    No significant hazards consideration comments received: No.

Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek 
Generating Station, Salem County, New Jersey

    Date of application for amendment: March 15, 2000.
    Brief description of amendment: This amendment changes Technical 
Specification (TS) Definition 1.7, CORE ALTERATION. The definition has 
been revised to be similar to the definition of CORE ALTERATION that is 
documented in NUREG-1433, Revision 1, ``Standard Technical 
Specifications, General Electric Plants, BWR/4.''
    Date of issuance: April 25, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 3 days.
    Amendment No.: 125.
    Facility Operating License No. NPF-57: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: March 23, 2000 (65 FR 
15657).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 25, 2000.
    No significant hazards consideration comments received: No.

Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek 
Generating Station, Salem County, New Jersey

    Date of application for amendment: September 30, 1999, as 
supplemented March 27, 2000.
    Brief description of amendment: This amendment revised the 
Technical Specifications (TSs) associated with the Safety Limit Minimum 
Critical Power Ratios in order to support the operation of Hope Creek 
Generating Station in the upcoming Cycle 10 with a mixed core of 
General Electric (GE) and Asea Brown Bovieri/Combustion Engineering 
(ABB/CE) fuel. In addition, administrative changes have been made to 
the TSs to reflect the change in fuel vendor from GE to ABB/CE.
    Date of issuance: May 1, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days after completion of Cycle 9.
    Amendment No.: 126.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 3, 1999 (64 FR 
59805).
    The March 27, 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 May 1, 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: July 20, 1999, as supplemented 
on October 25, 1999.
    Brief description of amendment: The amendment revises Technical 
Specifications to reflect the implementation of increased core flow.
    Date of Issuance: April 25, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 187.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46450). The October 25, 1999, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination or expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated April 25, 2000.
    No significant hazards consideration comments received: No.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: March 2, 2000.
    Brief description of amendment: The amendment increases the minimum 
refueling boron concentration to 2200 parts per million (ppm) from 2100 
ppm as specified in the Technical Specification 3.8.a.5.
    Date of issuance: May 1, 2000.
    Effective date: Immediately upon its date of issuance and is to be 
implemented within 30 days of the date of issuance.
    Amendment No.: 147.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.

Date of initial notice in Federal Register: March 30, 2000 (65 FR 
16969).

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 1, 2000.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of no Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of

[[Page 31367]]

increase in power output up to the plant's licensed power level, the 
Commission may not have had an opportunity to provide for public 
comment on its no significant hazards consideration determination. In 
such case, the license amendment has been issued without opportunity 
for comment. If there has been some time for public comment but less 
than 30 days, the Commission may provide an opportunity for public 
comment. If comments have been requested, it is so stated. In either 
event, the State has been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room).
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By June 16, 2000, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC and electronically from the ADAMS Public Library 
component on the NRC Web site, http://www.nrc.gov (the Electronic 
Reading Room). If a request for a hearing or petition for leave to 
intervene is filed by the above date, the Commission or an Atomic 
Safety and Licensing Board, designated by the Commission or by the 
Chairman of the Atomic Safety and Licensing Board Panel, will rule on 
the request and/or petition; and the Secretary or the designated Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

[[Page 31368]]

Duke Energy Corporation, et al., Docket No. 50-414, Catawba Nuclear 
Station, Unit 2, York County, South Carolina

    Date of application for amendments: May 3, 2000.
    Brief description of amendments: The amendment revised the 
Technical Specifications (TS) and associated Bases Section 3.6.9 for 
the Hydrogen Ignition System. Specifically, the proposed amendment 
modifies Surveillance Requirements (SRs) 3.6.9.1, 3.6.9.2, and 3.6.9.3 
to exclude the two hydrogen ignitors located beneath the reactor vessel 
missile shield from the applicability of the SRs. These two ignitors 
are presently considered to be inoperable at Unit 2 and cannot be 
accessed for replacement with the unit in its current operating mode 
(Mode 1). This change is effective for Unit 2 Cycle 11 only, or until 
such time that the unit enters Mode 5 (cold shutdown) such that the 
inoperable ignitors can be accessed for replacement.
    Date of issuance: May 5, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 178.
    Facility Operating License No. NPF-52: Amendment revised the 
Technical Specifications and associated Bases.
    Public Comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, consultation with the State of South Carolina, 
and final no significant hazards consideration determination are 
contained in a Safety Evaluation dated May 5, 2000.
    Attorney for licensee: Ms. Lisa F. Vaughn , Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: Richard L. Emch, Jr.

    Dated at Rockville, Maryland, this 10th day of May 2000.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-12302 Filed 5-16-00; 8:45 am]
BILLING CODE 7590-01-P