[Federal Register Volume 65, Number 230 (Wednesday, November 29, 2000)]
[Notices]
[Pages 71131-71144]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-30282]


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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 November 6, 2000, through November 16, 2000. 
The last biweekly notice was published on November 15, 2000.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC's Public Document 
Room, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. The filing of requests for a hearing 
and petitions for leave to intervene is discussed below.
    By December 29, 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, located at One 
White Flint North, 11555 Rockville Pike (first Floor), Rockville, 
Maryland 20852. Publicly available records will be accessible 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

[[Page 71132]]

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, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, 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, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 
Publicly available records will be accessible 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 amendment request: October 6, 2000 (U-603329).
    Description of amendment request: The proposed amendment would 
relocate Technical Specification Figure 3.6.4.1-1, ``Secondary 
Containment Drawdown Time for 1500 cfm Boundary Leakage'' to plant 
procedures.
    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 any accident previously 
evaluated.
    The proposed Technical specification (TS) change eliminates an 
inconsistency between the Secondary Containment surveillance 
requirement (SR 3.6.4.1.4) and the associated Bases. The proposed 
change (1) revises the wording of SR 3.6.4.1.4 to verify the time to 
draw down the secondary containment to 0.25 inch water 
gauge for each standby gas treatment (SGT) subsystem is within the 
required time; and (2) relocates the specific acceptance criteria 
(existing TS Figure 3.6.4.1-1) to plant procedures and the TS Bases.
    The scope of the proposed change is thus limited to the affected 
SR. No changes to plant equipment or the plant design are involved. 
The affected SR, as are surveillances in general, is not an 
initiator to any accident previously evaluated. Consequently, the 
probability of an accident previously evaluated is not significantly 
increased.
    The proposed change impacts SR 3.6.4.1.4 but does not change its 
intent or the associated acceptance criteria. Thus, the components 
and structural integrity being tested will still be required to be 
maintained Operable and capable of performing the accident 
mitigation functions assumed in the accident analysis. As a result, 
the consequences of any accident previously evaluated are not 
significantly affected. Therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) The proposed change would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed Technical Specification (TS) change eliminates an 
inconsistency between the Secondary Containment surveillance 
requirement (SR 3.6.4.1.4) and the associated Bases. The proposed 
change does not involve a physical alteration of the plant (no new 
or different type of equipment will be installed) or a change in the 
methods governing normal plant operation. No new failure modes are 
thus introduced by the proposed change. Therefore, this change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    (3) The proposed change will not involve a significant reduction 
in the margin of safety.
    The proposed Technical Specification (TS) change eliminates an 
inconsistency between the Secondary Containment Integrity 
surveillance requirement (SR 3.6.4.1.4) and the associated Bases. 
The revised wording of the Surveillance Requirement and the 
relocation of the acceptance criteria to plant procedures and TS 
Bases have been evaluated to ensure that they are sufficient to 
verify that the equipment used to meet the LCO can perform its 
required functions. The relocation of the acceptance criteria is 
consistent with the Bases previously approved in Amendment 21. Thus, 
appropriate equipment continues to be tested in a manner that gives 
confidence that the equipment can perform its assumed safety 
function. Therefore, this change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW, Washington, DC 20036-5869
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: October 6, 2000 (U-603332).
    Description of amendment request: The proposed amendment would 
remove from the Technical Specification surveillance requirements the 
minimum operating time specified

[[Page 71133]]

for the containment/hydrogen mixing system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    (1) The proposed change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    The proposed Technical Specification (TS) change deletes the 
minimum run time requirement (of 15 minutes) for the Containment/
Drywell Hydrogen Mixing System as denoted in Surveillance 
Requirement 3.6.3.3.1. Satisfying a TS Surveillance Requirement 
ensures that the associated system will function to mitigate the 
consequences of an accident, and, as such is not an initiator of an 
accident or malfunction. Therefore, since such a test requirement or 
operation is not an initiator to any accident previously evaluated, 
the probability of an accident previously evaluated is not 
significantly increased.
    In addition, the equipment being tested is still required to be 
operable and capable of performing its accident mitigation function 
assumed in the accident analysis. Eliminating the time requirement 
from the Surveillance Requirement does not reduce or relax the 
requirements to ensure that all controls are functioning properly. 
Also, it does not reduce or relax the requirements for ensuring the 
degraded conditions such as piping blockage, compressor failure or 
excessive vibration can be detected for corrective action. As a 
result, the consequences of any accident previously evaluated are 
not significantly affected. Therefore, this change does not involve 
a significant increase in the consequences of an accident previously 
evaluated.
    (2) The proposed change would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS change, as denoted in Surveillance Requirement 
3.6.3.3.1, does not involve a physical alteration of the plant (no 
new or different type of equipment will be installed) or a change in 
the methods governing normal plant operation. No new potential 
accident initiators are therefore introduced by the proposed change. 
Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) The proposed change will not involve a significant reduction 
in the margin of safety.
    The proposed TS change to delete the minimum run time 
requirement in Surveillance Requirement 3.6.3.3.1 does not result in 
a significant reduction in the margin of safety. As provided in the 
justification for the proposed change, the 15-minute run time 
acceptance criterion is not necessary to ensure that the 
Containment/Drywell Hydrogen Mixing can perform its required 
function. Thus, if a margin of safety can be ascribed to proper 
operation of this system for LOCA mitigation, the system will 
continue to be tested in a manner that gives confidence that it can 
perform its assumed safety function. Therefore, this change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW, Washington, DC 20036-5869.
    NRC Section Chief: Anthony J. Mendiola.

AmerGen Energy Company, LLC, Inc. et al., Docket No. 50-219, Oyster 
Creek Nuclear Generating Station, Ocean County, New Jersey

    Date of amendment request: August 29, 2000.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications in the ``Administrative Controls'' 
section to certain position titles and the Shift Technical Advisor 
(STA) staffing requirement to allow one of the required on-shift Senior 
Reactor Operator (SRO) positions to be combined with the required STA 
position so as to serve in a dual role SRO/STA position.
    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. Would operation of the facility in accordance with the 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    The [proposed] changes do not affect the purpose, function, 
performance, operability, inspection or testing of and does not make 
any physical or procedural changes to plant systems, structures or 
components. Also, all existing technical specification limiting 
conditions for operation and surveillance requirements are retained.
    TSCR [Technical Specifications Change Request] makes 
administrative changes to certain position titles without changing 
the technical requirements for position responsibilities.
    TSCR also changes the Shift Technical Advisor (STA) staffing 
requirement to allow one of the required on-shift Senior Reactor 
Operator (SRO) positions to be combined with the required STA 
position so as to serve in a dual-role SRO/STA position as 
encouraged by the NRC in Option 1 of Generic Letter 86-04, ``Policy 
Statement on Engineering Expertise On Shift'', dated February 13, 
1986.
    Implementation of the proposed dual-role SRO/STA change will 
result in personnel with enchanced operational knowledge being 
assigned to perform the STA function of providing accident 
assessment expertise, and analyzing and responding to off normal 
occurrences when needed. The NRC's stated preference in the October 
28, 1985, ``Policy Statement on Engineering Expertise on Shift'', 
indicates that the NRC has concluded that the individual filling the 
dual-role SRO/STA position may perform these functions better than a 
non-licensed individual filling the STA position even when the SRO/
STA is concurrently functioning as one of the required shift SROs.
    Therefore, since no physical or procedural changes are being 
made to existing plant systems, structures or components and since 
the position title changes are administrative in nature and the 
function and responsibilities of the STA will be executed by an 
appropriately qualified individual filling the dual-role SRO/STA 
position, operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Would operation of the facility in accordance with the 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    The [proposed] changes do not affect the purpose, function, 
performance, operability, inspection or testing of and does not make 
any physical or procedural changes to plant systems, structures or 
components. Also, all existing technical specification limiting 
conditions for operation and surveillance requirements are retained.
    TSCR 277 makes administrative changes to certain position titles 
without changing the technical requirements for the position 
responsibilities.
    TSCR 277 also changes the Shift Technical Advisor (STA) staffing 
requirement to allow one of the required on-shift Senior Reactor 
Operator (SRO) positions to be combined with the required STA 
position so as to serve in a dual-role SRO/STA position as 
encouraged by the NRC in Option 1 of Generic Letter 86-04, ``Policy 
Statement on Engineering Expertise On Shift'', dated February 13, 
1986.
    Therefore, since no physical or procedural changes are being 
made to existing plant systems, structures or components and since 
the position title changes are administrative in nature and the 
function and responsibilities of the STA will be executed by an 
appropriately qualified individual filling the dual-role SRO/STA 
position, operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Would operation of the facility in accordance with the 
proposed change involve a significant reduction in a margin of 
safety?
    The [proposed] changes do not affect the purpose, function, 
performance, operability, inspection or testing of and does not make

[[Page 71134]]

any physical or procedural changes to plant systems, structures or 
components. Also, all existing technical specification limiting 
conditions for operation and surveillance requirements are retained.
    TSCR 277 makes administrative changes to certain position titles 
without changing the technical requirements for the position 
responsibilities.
    TSCR 277 also changes the Shift Technical Advisor (STA) staffing 
requirement to allow one of the required on-shift Senior Reactor 
Operator (SRO) positions to be combined with the required STA 
position so as to serve in a dual-role SRO/STA position as 
encouraged by the NRC in Option 1 of Generic Letter 86-04, ``Policy 
Statement on Engineering Expertise On Shift'', dated February 13, 
1986.
    Therefore, since no physical or procedural changes are being 
made to existing plant systems, structures or components and since 
the position title changes are administrative in nature and the 
function and responsibilities of the STA will be executed by an 
appropriately qualified individual filling the dual-role SRO/STA 
position and shift staffing required by TS 6.2.2.2 and 10 CFR 
50.54(m)(2) will [continue] to be maintained, operation of the 
facility in accordance with the proposed amendment will not involve 
a significant reduction in [a] margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Jr., Esquire, Morgan, 
Lewis, & Bockius LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: M. Gamberoni.

Consolidated Edison Company of New York, Inc., Docket No. 50-003, 
Indian Point Nuclear Generating Station, Unit 1, Buchanan, New York

    Date of amendment request: October 5, 2000.
    Brief description of amendment: The proposed changes would revise 
Technical Specifications (TSs) Sections 3.2.1.a, 3.2.1.e, and 3.2.1.f 
to relocate administrative controls to the Quality Assurance Program 
Description (QAPD).
    Basis for proposed no significant hazards considerations 
determination: As required by 10 CFR 50.91(a), the licensee has provide 
its analysis of the issue of no significant hazards consideration, 
which is presented below:

    (1) Does the proposed license amendment involve a significant 
increase in the probability or in the consequences of an accident 
previously evaluated?
    No. The proposed changes are administrative in nature. The 
changes involve Section 3.2.1.a, referencing the QAPD instead of the 
IP#2 [Indian Point Nuclear Generating Station, Unit 2] UFSAR 
[Updated Final Safety Analysis Report], deleting Section 3.2.1.e and 
having Section 3.2.1.f, refer to the QAPD for the administrative 
controls. These changes do not affect possible initiating events for 
accidents previously evaluated or alter the configuration or 
operation of the facility. The Limiting Safety System Settings and 
Safety Limits specified in the current Technical Specifications 
remain unchanged. Therefore, the proposed changes would not involve 
a significant increase in the probability or in the consequences of 
an accident previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed changes are administrative in nature. The 
safety analysis of the facility remains complete and accurate. There 
are no physical changes to the facility and the plant conditions for 
which the design basis accidents have been evaluated are still 
valid. The operating procedures and emergency procedures are 
unaffected. Consequently no new failure modes are introduced as a 
result of the proposed changes. Therefore, the proposed changes 
would not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    No. The proposed changes are administrative in nature. Since 
there are no changes to the operation of the facility or the 
physical design, the IP#1 [Indian Point Nuclear Generating Station, 
Unit 1] FSAR [Final Safety Analysis Report] or the IP#2 UFSAR design 
basis, accident assumptions, or IP#1 Technical Specification Bases 
are not affected. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards considerations.
    Attorney for Licensee: Brent L. Brandenberg, Esq., Assistant 
General Counsel, Consolidated Edison Company of New York, Inc., 4 
Irving Place-1830, New York, NY 10003.
    NRC Section Chief: Michael T. Masnik.

Energy Northwest, Docket No. 50-397, WNP-2, Benton County, Washington

    Date of amendment request: October 12, 2000.
    Description of amendment request: The amendment changes the 
facility name of WNP-2 to Columbia Generating Station in all the 
applicable portions of the Operating License including Appendix A 
(Technical Specifications) and Appendix B (Environmental Protection 
Plan). In addition, the proposed change makes editorial changes to 
Technical Specification Figure 4.1-1, Site Area Boundary.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    No. This request involves an administrative change only. The 
Operating License (OL) and Technical Specification Figure 4.1-1, 
Site Area Boundary, are being changed to reflect the new name of the 
facility. In addition, editorial changes are being made to Figure 
4.1-1 for clarification. No actual plant equipment or accident 
analyses are affected by the proposed change. Therefore, this 
request will have no impact on the probability or consequence of any 
type of 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.
    No. This request involves an administrative change only. The OL 
and Technical Specification Figure 4.1-1 are being changed to 
reflect the new name of the facility. In addition, editorial changes 
are being made to Figure 4.1-1 for clarification. No actual plant 
equipment or accident analyses are affected by the proposed change 
and no failure modes not bounded by previously evaluated accidents 
will be created. Therefore, this request will have no impact on the 
possibility of any type of accident: new, different, or previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    No. Margin of safety is associated with confidence in the 
ability of the fission product barriers (i.e., fuel and fuel 
cladding, Reactor Coolant System pressure boundary, and containment 
structure) to limit the level of radiation dose to the public. This 
request involves an administrative change only. The OL and Technical 
Specification Figure 4.1-1 are being changed to reflect the new name 
of the facility. In addition, editorial changes are being made to 
Figure 4.1-1 for clarification.
    No actual plant equipment or accident analyses are affected by 
the proposed change. Additionally, the proposed change will not 
relax any criteria used to establish safety limits, will not relax 
any safety system settings, or will not relax the bases for any 
limiting conditions of operation. Therefore, this proposed change 
will not impact 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

[[Page 71135]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

Energy Northwest, Docket No. 50-397, WNP-2, Benton County, Washington

    Date of amendment request: October 30, 2000.
    Description of amendment request: Surveillance Requirement (SR) 
3.6.1.3.8 currently requires verification of the actuation capability 
of each excess flow check valve (EFCV) every 24 months. The proposed 
change is to relax the SR frequency by allowing a ``representative 
sample'' of reactor instrument line EFCVs to be tested every 24 months, 
such that each reactor instrument line EFCV will be tested at least 
once every 10 years (nominal). The proposed change will also result in 
limiting the SR to only the reactor instrument line EFCVs.
    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 consequence of an accident previously 
evaluated.
    The current SR frequency requires each reactor instrument line 
EFCV to be tested every 24-months. The reactor instrument line EFCVs 
at WNP-2 are designed so that they will not close accidentally 
during normal operation, but will close if a rupture of the 
instrument line is indicated downstream of the valve, and have their 
status indicated in the control room. This proposed change allows a 
reduced number of reactor instrument line EFCVs to be tested every 
24-months. There are no physical plant modifications associated with 
this change. Industry operating experience demonstrates a high 
reliability of these valves. Neither reactor instrument line EFCVs 
nor their failures are capable of initiating previously evaluated 
accidents; therefore, there can be no increase in the probability of 
occurrence of an accident regarding this proposed change.
    Reactor instrument lines connecting to the reactor coolant 
pressure boundary are equipped with EFCVs and also have a flow-
restricting orifice inside containment and upstream of the EFCV. The 
consequences of an unisolable rupture of such an instrument line has 
been previously evaluated in WNP-2 FSAR [Final Safety Analysis 
Report] 15.6.2. The instrument lines that penetrate primary 
containment conform to Regulatory Guide 1.11 (WNP-2 FSAR 7.1.2.4). 
Those instrument lines are Seismic Category I and terminate in 
instruments that are Seismic Category I (reference WNP-2 FSAR Table 
6.2-16 note 27).
    The sequence of events in WNP-2 FSAR Section 15.6.2.2 for a 
reactor instrument line break assumes a continuous discharge of 
reactor water through the instrument line until the reactor vessel 
is cooled and depressurized (5 hours). Although not expected to 
occur as a result of this change, the postulated failure of an EFCV 
to isolate as a result of reduced testing is bounded by this 
previous evaluation. Therefore, there is no increase in the 
previously evaluated consequences of the rupture of an instrument 
line and there is no potential increase in the consequences of an 
accident previously evaluated as a result of this change.
    The containment atmosphere and suppression pool instrument line 
EFCVs are required to remain open to sense containment atmosphere 
and suppression pool level conditions during postulated accidents. 
They are not required to close during an instrument line break 
assumed during normal plant operation nor is their design capable of 
closing during normal plant conditions. These EFCVs do not meet the 
criteria for inclusion in 10 CFR 50.36(c)(3) as they have no active 
safety function and thus relocation of their testing requirements to 
the FSAR cannot effect the probability of an increase in the 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This proposed change allows a reduced number of reactor 
instrument line EFCVs to be tested each operating cycle and that the 
testing requirements for containment atmosphere and suppression pool 
instrument line EFCVs be relocated to the FSAR. No other changes in 
requirements are being proposed. Industry operating experience 
demonstrates the high reliability of these valves. The potential 
failure of a reactor instrument line EFCV to isolate by the proposed 
change in testing is bounded by the previous evaluation of an 
instrument line rupture. This change will not physically alter the 
plant (no new or different type of equipment will be installed). 
This change will not alter the operation of process variables, 
structures, or components as described in the safety analysis. Thus, 
a new or different kind of accident will not be created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The consequences of an unisolable rupture of an instrument line 
has been evaluated in WNP-2 FSAR Section 15.6.2 in accordance with 
the requirements of Regulatory Guide 1.11. That evaluation assumed a 
continuous discharge of reactor water for the duration of the 
detection and cooldown sequence (5 hours). The only margin of safety 
applicable to this proposed change is considered to be that implied 
by this evaluation. Since a continuous discharge was assumed in this 
evaluation, any potential failure of a reactor instrument line EFCV 
to isolate as a result of reduced testing frequency is bounded by 
existing analysis and does not involve a significant reduction in 
the margin of safety.
    There is no accident for which the containment atmosphere or 
suppression pool instrument line EFCVs are designed to actuate to 
the isolation position for mitigation. A postulated break of a 
containment atmosphere or suppression pool instrument line under 
normal operating conditions would not result in a condition that 
would create the ability for these EFCVs to operate because neither 
the containment pressure nor the suppression pool level head would 
be sufficient to result in their actuation. As these EFCVs have no 
active design or safety function, the relocation of testing 
requirements would not involve a significant reduction in the margin 
of safety. A postulated break of any instrument line simultaneously 
with a loss of coolant accident is beyond the design basis for the 
plant.

    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: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

GPU Nuclear Corporation and Saxton Nuclear Experimental Corporation 
(SNEC), Docket No. 50-146, Saxton Nuclear Experimental Facility (SNEF), 
Bedford County, Pennsylvania

    Date of amendment request: February 2, 2000, as supplemented on 
August 11 and September 18, 2000.
    Description of amendment request: The proposed amendment would 
approve the license termination plan for the SNEF.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed change is necessary to achieve the decommissioning 
objective of terminating the license and releasing the site for 
unrestricted use. As such, the proposed change:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated since accidents 
which might occur during the active decommissioning phase of the 
SNEC facility are bounded by the twelve accidents addressed in 
section 3.0 of the Updated Safety Analysis Report (USAR). The 
accident analysis addressed in the USAR demonstrate that no adverse 
public health and safety impacts are expected from accidents that

[[Page 71136]]

might occur during decommissioning operations at the SNEC facility. 
The greater part of radioactively contaminated materials and 
components originally located in the SNEC facility Containment 
Vessel are no longer on site, having been shipped as radioactive 
waste.
    Implementation of the SNEC License Termination Plan involves a 
continuation of the decommissioning process including the final 
status survey activity to be performed prior to site closeout at the 
end of the dismantlement phase. These activities do not involve a 
significant increase in either the probability or consequences of an 
accident previously evaluated.
    2. Will not create the possibility of a new or different kind of 
accident from any accident previously evaluated. Accidents 
previously evaluated in the USAR access different methods of 
dispersing radioactive material to the environment, which include a 
loss of support systems and external events. Remaining dismantlement 
activities and final status survey work described in the License 
Termination Plan are similar to those previously performed and will 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Will not involve a significant reduction in a margin of 
safety. The Technical Specifications currently in place at the SNEC 
facility were developed to safely decommission the SNEC facility. 
Issuance of the proposed amendment would not reduce the controls 
established by the technical specifications for activities performed 
at the SNEC facility. The proposed License Amendment establishes 
additional controls to ensure License Termination Plan activities 
are performed effectively. Thus, this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the analysis of the licensees 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 the Licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts, and Trowbridge, 2300 N Street, N.W., Washington, DC 
20037.
    NRC Branch Chief: Ledyard B. Marsh.

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

    Date of amendment request: June 1, 2000.
    Description of amendment request: The licensee is proposing to 
relocate Technical Specifications 3.3.3.2, ``Instrumentation, Movable 
Incore Detectors''; 3.3.3.3, ``Instrumentation, Seismic 
Instrumentation''; 3.3.3.4, ``Instrumentation, Meteorological 
Instrumentation''; 3.3.3.8, ``Loose-Part Detection System''; and 3.3.4, 
``Turbine Overspeed Protection'' and Index Pages vi and vii to the 
Technical Requirements Manual (TRM). The Bases of the affected 
Technical Specifications will be modified to address the proposed 
changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to relocate the movable incore detector 
instrumentation, seismic monitoring instrumentation, meteorological 
monitoring instrumentation, loose-part detection instrumentation, 
and turbine overspeed protection instrumentation from the Technical 
Specifications to the TRM will have no adverse effect on plant 
operation, or the availability or operation of any accident 
mitigation equipment. The plant response to the [design-basis 
accidents] DBAs will not change. In addition, the movable incore 
detector instrumentation, seismic monitoring instrumentation, 
meteorological monitoring instrumentation, and loose-part detection 
instrumentation are not accident initiators and cannot cause an 
accident. For the turbine overspeed protection instrumentation, the 
DBAs and transients include a variety of system failures and 
conditions which might result from turbine overspeed events and 
potential missiles striking various plant systems and equipment. 
However, in view of the low likelihood of the generation of turbine 
missiles, the turbine overspeed protection instrumentation does not 
serve a primary protective function. Therefore, these changes will 
not significantly increase 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 to relocate the movable incore detector 
instrumentation, seismic monitoring instrumentation, meteorological 
monitoring instrumentation, loose-part detection instrumentation, 
and turbine overspeed protection instrumentation do not alter the 
plant configuration (no new or different type of equipment will be 
installed) or require any new or unusual operator actions. They do 
not alter the way any [structure, system, or component] SSC 
functions and do not alter the manner in which the plant is 
operated. The proposed changes do not introduce any new failure 
modes. Therefore, the proposed changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed Technical Specification and Bases changes will 
relocate the requirements for the movable incore detector 
instrumentation, seismic monitoring instrumentation, meteorological 
monitoring instrumentation, loose-part detection instrumentation, 
and turbine overspeed protection instrumentation from Technical 
Specifications to the TRM. Any future changes to the relocated 
requirements will be in accordance with 10 CFR 50.59 and approved 
station procedures. The proposed changes will have no adverse effect 
on plant operation, or the availability or operation of any accident 
mitigation equipment. The plant response to the DBAs will not 
change. In addition, the relocated requirements do not meet any of 
the 10 CFR 50.36c(2)(ii) criteria on items for which Technical 
Specifications must be established. Therefore, there will be no 
significant reduction in a margin of safety.

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

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

    Date of amendment request: June 29, 2000 as supplemented on October 
16, 2000.
    Description of amendment request: The proposed changes would modify 
Technical Specification (TS) Sections 3.3.2, ``Instrumentation--
Engineered Safety Features Actuation System Instrumentation;'' 3.7.7, 
``Plant Systems--Control Room Emergency Ventilation System;'' 3.7.8, 
``Plant Systems--Control Room Envelope Pressurization System;'' 3.7.9, 
``Plant Systems--Auxiliary Building Filter System;'' 3.9.1.1, 
``Refueling Operations--Boron Concentration,'' 3.9.1.2; ``Refueling 
Operations--Boron Concentration;'' 3.9.2, ``Refueling Operations--
Instrumentation;'' 3.9.4, ``Refueling Operations--Containment Building 
Penetrations;'' 3.9.9, ``Refueling Operations--Containment Purge and 
Exhaust Isolation System;'' 3.9.10, ``Refueling Operations--Water 
Level--Reactor Vessel;'' and 3.9.12, ``Refueling Operations--Fuel 
Building Exhaust Filter System.'' Some of these proposed changes are 
associated with the revised fuel handling accident analysis, and 
integrity of the Control Room and the Fuel Building boundaries. Several 
administrative changes are also proposed to reflect Millstone Unit 3 
terminology, removal of unnecessary information and to eliminate 
confusion

[[Page 71137]]

by providing consistency between limiting conditions for operation, 
action requirements, and Surveillance Requirements. The proposed 
Technical Specifications changes associated with the revised 
containment fuel handling accident analysis results in an increase in 
the consequences of a containment fuel handling accident since the 
current analysis of a containment fuel handling accident does not 
assume the release of any radioactive material from containment. The 
revised analysis assumes a release of radioactive material because it 
assumes both personnel access hatch doors are open and at least one 
hatch door is closed within 10 minutes of a fuel handling accident 
inside containment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. 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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.

Technical Specification Changes Associated with Analyses Changes

    The proposed Technical Specification changes associated with the 
revised fuel handling accident analyses will not cause an accident 
to occur and will not result in any change in the operation of the 
associated accident mitigation equipment. The design basis accidents 
remain the same postulated events described in the Millstone Unit 
No. 3 FSAR. Therefore, the proposed changes will not increase the 
probability of an accident previously evaluated.
    The proposed Technical Specification changes associated with the 
revised fuel handling accidents analyses will increase the 
associated consequences. The increased consequences are the result 
of a revised plant configuration and revised calculation 
assumptions, not the result of the addition of any new plant 
equipment. The current Fuel Handling Accident Inside Containment 
(FHAIC) analysis assumes the containment is isolated, or will be 
isolated, prior to any release. The revised FHAIC analysis will 
allow both containment personnel access hatch doors to remain open, 
under administrative control, during core alterations and irradiated 
fuel movement inside containment. This may result in a radioactive 
release if a fuel handling accident were to occur. The revised FHAIC 
analysis demonstrates that the magnitude of the potential release is 
small and bounded by the consequences of the Design Basis Loss of 
Coolant Accident. The increase in the consequences of the revised 
Fuel Handling Accident Inside the Spent Fuel Pool (FHAISFP) analysis 
due to the revised calculation assumptions is small. Therefore, the 
proposed changes will not result in a significant increase in the 
consequences of an accident previously evaluated.

Other Technical Specification Changes

    The proposed Technical Specification changes not associated with 
the revised fuel handling accidents analyses affect the limiting 
conditions for operation (LCOs), applicability, action requirements, 
and surveillance requirements of numerous specifications associated 
with plant operating restrictions, accident mitigation functions, 
and accident mitigation equipment. The affected operating 
restrictions, accident mitigation functions, and accident mitigation 
equipment are not accident initiators. The proposed changes will not 
cause an accident to occur and will not result in any change in the 
operation of the associated accident mitigation equipment. The 
design basis accidents remain the same postulated events described 
in the Millstone Unit No. 3 FSAR. Therefore, the proposed changes 
will not increase the probability of an accident previously 
evaluated.
    The proposed LCO and applicability changes are consistent with 
the design basis accident analyses, including the revised fuel 
handling accident analyses. (The proposed change to the LCO for 
containment penetrations, which will allow both personnel access 
hatch doors to remain open during core alterations and irradiated 
fuel movement inside containment will result in an increase in the 
consequences of a FHAIC as previously discussed.) This will ensure 
that the accident mitigation functions and associated equipment are 
available for accident mitigation as assumed in the associated 
analyses. As a result, the accident analysis assumptions and 
mitigation methods will not be adversely affected by these changes. 
Therefore, the proposed changes will not result in a significant 
increase in the consequences of an accident previously evaluated.
    The additional proposed changes to the Technical Specifications 
that will standardize terminology, relocate information to the 
Bases, remove extraneous information, and make minor format changes 
will not result in any technical changes to the current 
requirements. Therefore, these additional proposed changes will not 
result in 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 to the Technical Specifications do not 
impact any system or component that could cause an accident. The 
proposed changes will not alter the plant configuration (no new or 
different type of equipment will be installed) or require any 
unusual operator actions. The proposed changes will not alter the 
way any structure, system, or component functions, and will not 
significantly alter the manner in which the plant is operated. There 
will be no adverse effect on plant operation or accident mitigation 
equipment. The response of the plant and the operators following an 
accident will not be significantly different. In addition, the 
proposed changes do not introduce any new failure modes. Therefore, 
the proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    Title 10 of the Code of Federal Regulations Part 100 establishes 
the accident exposure limits (300 rem thyroid and 25 rem whole body) 
for the Exclusion Area Boundary and Low Population Zone. The 
radiological consequences resulting from the Technical Specification 
changes associated with the revised fuel handling accident analyses 
are well within these limits. The radiological consequences to the 
Control Room Operators resulting from the Technical Specification 
changes associated with the revised fuel handling accident analyses 
are also within the GDC 19 limit. Since these limits will not be 
exceeded and these limits establish the margin of safety in the 
plant's current licensing basis, the proposed changes will not 
result in a significant reduction in a margin of safety.
    The proposed Technical Specification LCO, applicability, action 
requirement, and surveillance requirement changes not associated 
with the revised fuel handling accidents analyses do not adversely 
affect equipment design or operation. In addition, the proposed 
allowed outage times and shutdown times are consistent with times 
already contained in the Millstone Unit No. 3 Technical 
Specifications. Therefore, these changes will not result in a 
significant reduction in a margin of safety.
    The additional proposed changes to the Technical Specifications 
that will standardize terminology, relocate information to the 
Bases, remove extraneous information, and make minor format changes 
will not result in any technical changes to the current 
requirements. Therefore, these additional changes will not result in 
a significant reduction in a margin of safety.
    Based on the staff's analysis, it appears that the three standards 
of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: James W. Clifford.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: October 12, 2000.
    Description of amendment request: The proposed amendment would make 
changes to Hope Creek Generating Station (HCGS) Technical 
Specifications (TS) and Bases associated with the drywell vacuum 
breakers and the suppression pool vacuum breakers. The proposed changes 
are intended to provide consistency between the HCGS TS and the 
Standard Technical

[[Page 71138]]

Specifications (STS) (NUREG-1433). These changes include revising or 
deleting specific Limiting Conditions for Operation and Surveillance 
Requirements and include relocating information from these sections to 
the Bases. In addition, a change to the Containment Systems 
Surveillance Requirements was proposed to correct the hierarchical 
format of that section.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is 
presented below:
    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The function of the suppression-chamber-to-drywell vacuum 
breakers is to relieve vacuum in the drywell by allowing air and 
steam flow from the suppression chamber to the drywell when the 
drywell is at a negative pressure with respect to the suppression 
chamber. A negative differential pressure across the drywell wall is 
caused by rapid depressurization of the drywell. Steam condensing in 
the drywell as a result of a primary system rupture results in the 
most severe pressure transient.
    The function of the reactor building-to-suppression chamber 
vacuum breakers is to relieve vacuum when the primary containment 
depressurizes below the reactor building pressure. If the drywell 
depressurizes below reactor building pressure, the negative 
differential pressure is mitigated by flow through the reactor 
building-to-suppression chamber vacuum breakers and through the 
suppression-chamber-to-drywell vacuum breakers. A negative 
differential pressure across the drywell wall is caused by rapid 
depressurization of the drywell. The maximum depressurization rate 
is a function of the primary containment spray flow rate and 
temperature and the assumed initial conditions of the primary 
containment atmosphere.
    Each proposed change to the vacuum breaker TS was categorized by 
the licensee as either administrative, more restrictive, less 
restrictive, or as a relocation. In addition, the licensee made 
changes to the bases to capture the information removed from the 
associated Action Steps and to be consistent with the STS. The 
administrative changes eliminate, replace, or add words or phrases, 
to provide clarity or to achieve consistency with the STS. The more 
restrictive changes reduce the number of vacuum breakers allowed to 
be open or reduces the amount of time allowed to close the open 
valves. The less restrictive changes: (1) Eliminate the surveillance 
requirements associated with the vacuum breaker position indicators; 
(2) reduce the frequency of vacuum breaker position verification; 
(3) increase the time requirement for functional testing subsequent 
to steam discharged to the suppression chamber from the safety-
relief valves; (4) increase the number of allowable inoperable 
valves in one vacuum breaker assembly; (5) eliminate repetitious 
visual inspections; and (6) eliminate channel calibration as a means 
to determine operability of the inboard isolation valve auto-open 
control system. The relocation changes move information from the 
action steps to the bases.
    The licensee stated in their October 12, 2000, application that 
neither the vacuum breakers, the vacuum breaker position indication, 
nor the vacuum breaker actuation system are initiators of any 
analyzed event. Therefore, the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated. In addition, the licensee's application states that the 
proposed changes will ensure the operability of the vacuum breakers, 
will provide assurance that the containment integrity and venting 
capability are maintained or restored within 1 hour, and that 
sufficient vacuum breakers will remain operable to mitigate the 
assumed accidents. Therefore, there is no significant increase in 
the consequences of an accident previously evaluated. The licensee 
further stated that any future changes to the licensee-controlled 
documents containing relocated requirements will be evaluated in 
accordance with the PSEG Nuclear 10 CFR 50.59 program. Consequently, 
no significant increase in the consequences of an accident 
previously evaluated will be allowed without prior Nuclear 
Regulatory Commission (NRC) approval. Therefore, the proposed change 
does not involve a significant increase in the consequences of an 
accident previously evaluated.
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously analyzed.
    The licensee stated in their application that the proposed 
change does not involve a physical alteration of the plant and does 
not introduce any new modes of plant operation. In addition, the 
licensee stated that any resulting changes to the operation of the 
plant will be consistent with assumptions made in the safety 
analysis. Therefore the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously analyzed.
    The proposed change does not involve a significant reduction in 
a margin of safety.
    The licensee stated in their application that the proposed 
change does not impact any safety analysis assumptions and will 
provide assurance that the containment integrity and venting 
capability are maintained or restored within 1 hour. The licensee 
further stated that Hope Creek experience has shown that the change 
in surveillance frequency of vacuum breaker position is not a 
significant change in operating practice. In addition, the 
operability of the vacuum breakers is not adversely affected by 
steam discharged through the safety relief valves (SRVs) and does 
not pose an immediate operability concern. Consequently, the 
potential impact from the proposed increase in the amount of time 
during which to perform functional testing subsequent to an SRV lift 
is minimal. Therefore, there is no significant reduction in a margin 
of safety. The licensee further stated that any future changes to 
the licensee-controlled documents containing relocated requirements 
will be evaluated in accordance with the PSEG Nuclear 10 CFR 50.59 
program. Consequently, no significant reduction in a margin of 
safety will be allowed without prior NRC approval. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.
    Based on the staff's analysis, it appears that the three standards 
of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, PSEG Nuclear--
N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: October 30, 2000 (TS-407).
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to remove the term ``maximum 
pathway'' from the main steam isolation valve (MSIV) leakage rate 
Surveillance Requirement (SR) 3.6.1.3.10. This proposed change would 
provide consistency with 10 CFR Part 50 Appendix J leak rate testing 
terminology for evaluating MSIV leakage rates.
    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 eliminate the words ``maximum pathway'' 
does not affect any plant system or component, and does not impact 
operator performance or procedures. The leak rate testing of the 
MSIVs will continue to be performed in accordance with 10 CFR 50 
Appendix J in a manner consistent with the guidance on leak rate 
testing presented in industry guidance documents and in the Standard 
TS. The change does not impact the design basis accident analyses 
presented in the Final Safety Analysis Report (FSAR). This proposed 
TS change is considered administrative in that no changes in leak 
testing methods or in disposition of leak rate results are involved. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 71139]]

    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    No changes in accident analysis are involved, so the 
consequences of accidents will remain within the accident analysis 
described in the FSAR. Therefore, the proposed TS change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change does not affect any plant system or 
component, and does not have any impact on plant operation. No 
changes in accident analyses are involved, therefore, the proposed 
change does not involve a significant reduction in the margin of 
safety as currently defined in the bases of the applicable TS 
section or in the FSAR. For these reasons, the proposed amendment 
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 I0H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: November 6, 2000 (TS-411).
    Description of amendment request: The proposed amendment would 
revise the technical specifications to allow two Residual Heat Removal 
(RHR) suppression pool cooling subsystems to be inoperable for 8 hours.
    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 change does not result in any hardware or operating 
procedure changes. The RHR Suppression Pool Cooling subsystems are 
not assumed to be initiators of any analyzed event. This change 
allows an additional 8 hours to restore required RHR Suppression 
Pool Cooling subsystem(s) prior to requiring the initiation of a 
unit shutdown. The proposed 8 hour Completion Time provides some 
time to restore required subsystem(s) to Operable status, yet is 
short enough that operating an additional 8 hours is not a 
significant risk.
    Therefore, the proposed amendment 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 possibility of a new or different kind of accident from any 
previously evaluated is not created because the proposed change 
introduces no new mode of plant operation and it does not involve a 
physical modification to the plant.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The increased time allowed for restoring required inoperable RHR 
Suppression Pool Cooling subsystems is acceptable based on the small 
probability of an event requiring the inoperable suppression pool 
cooling subsystems to function and the desire to restore required 
subsystems prior to requiring the initiation of a plant shutdown. 
Delaying a plant shutdown will minimize the potential for a scram 
which then could result in a need for a subsystem when it is 
inoperable. As such, any reduction in a margin of safety will be 
insignificant and offset by the benefit gained from providing 
additional time to restore required subsystem(s), thus avoiding 
potential plant transients during shutdown. Therefore, the proposed 
change does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET I0H, Knoxville, Tennessee 37902
    NRC Section Chief: Richard P. Correia.

Three Mile Island Nuclear Station, Unit 2, Docket No. 50-320, 
Middletown, Pennsylvania

    Date of amendment request: August 9, 2000.
    Brief description of amendment request: The proposed technical 
specifications change request (TSCR) is to revise Three Mile Island 
Nuclear Generating Station, Unit 2 (TMI-2), Technical Specifications 
Sections 6.5.3.2, 6.5.4.1, 6.5.4.2.a, 6.5.4.2.b, 6.5.4.3, 6.5.4.3.c, 
6.5.4.4 and 6.5.4.6, to eliminate the reference to Independent Onsite 
Safety Review Group (IOSRG) and to define the performance of the IOSRG 
function by the nuclear quality assurance organization. Also, two 
titles that no longer exist (Manager, TMI-2 Department and division 
vice president) were corrected. These administrative changes are 
similar to changes that have been already approved at other plants in 
Region I.
    Basis for proposed no significant hazards consideration: As 
required by 10 CFR 50.91(a), the license has provided its analysis of 
the issue of no significant hazards consideration, which is presented 
below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated. The proposed changes do not affect assumptions 
contained in plant safety analyses, the physical design and/or 
operation of the plant, nor do they affect Technical Specifications 
that preserve safety analysis assumptions. None of the proposed 
changes involve a physical modification to the plant, a new mode of 
operation or a change to the UFSAR [Updated Final Safety Analysis 
Report] transient analyses. No Technical Specification Limiting 
Condition for Operation, Action Statement, or Surveillance 
Requirement is affected by any of the proposed changes. The proposed 
changes do not alter the design, function, or operation of any plant 
component. Therefore, the proposed amendment does not affect the 
probability of occurrence or consequences of an accident previously 
evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any previously evaluated. The proposed changes 
do not affect assumptions contained in plant safety analyses, the 
physical design and/or modes of plant operation defined in the plant 
operating license, or Technical Specifications that preserve safety 
analysis assumptions. The proposed changes do not introduce a new 
mode of plant operation or surveillance requirement, nor involve a 
physical modification to the plant. The proposed changes do not 
alter the design, function, or operation of any plant components. 
Therefore, the proposed amendment does not affect the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. None of the proposed changes involve a physical modification 
to the plant, a new mode of operation or a change to the UFSAR 
transient analyses. No Technical Specification Limiting Condition 
for Operation, Action Statement, or Surveillance Requirement is 
affected. Therefore, the proposed amendment does not reduce the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three

[[Page 71140]]

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. Esq., Shaw, Pittman, 
Potts & Trowbridge 2300 N. Street, N.W., Washington, DC 20037.
    NRC Section Chief: Mike Masnik.

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

    Date of amendment request: October 25, 2000.
    Description of amendment request: The proposed amendment would make 
editorial and administrative changes to the Technical Specifications 
(TS). These changes correct spelling and grammatical errors, correct 
references, eliminate excessive detail related to specifying a job 
title, revise position titles, consolidate pages and generalize 
statements allowing Nuclear Regulatory Commission (NRC) approved 
alternatives to specified requirements.
    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 operation of the Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes are administrative or editorial in nature 
and do not involve any physical changes to the plant. The changes do 
not revise the methods of plant operation which could increase the 
probability or consequences of accidents. No new modes of operation 
are introduced by the proposed changes such that a previously 
evaluated accident is more likely to occur or more adverse 
consequences would result.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    These changes are administrative or editorial in nature and do 
not affect the operation of any systems or equipment, nor do they 
involve any potential initiating events that would create any new or 
different kind of accident. There are no changes to the design 
assumptions, conditions, configuration of the facility, or manner in 
which the plant is operated and maintained.
    The changes do not affect assumptions contained in plant safety 
analyses or the physical design and/or modes of plant operation. 
Consequently, no new failure mode is introduced.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    There are no changes being made to the TS safety limits or 
safety system settings. The operating limits and functional 
capabilities of systems, structures and components are unchanged as 
a result of these administrative and editorial changes. These 
changes do not affect any equipment involved in potential initiating 
events or plant response to accidents. There is no change to the 
basis for any Technical Specification that is related to the 
establishment or maintenance of, a nuclear safety margin.

    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

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.

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

    Date of application for amendments: October 30, 2000.
    Brief description of amendments: Modify the Technical 
Specifications to allow a one-time-only increase in the diesel 
generator Action Completion Time from 72 hours to 10 days to facilitate 
potential repairs to an emergency diesel generator to improve 
reliability.
    Date of publication of individual notice in the Federal Register: 
November 3, 2000 (65 FR 66266).
    Expiration date of individual notice: December 4, 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, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. 
Publicly available records will be accessible electronically from the 
ADAMS Public Library component on the NRC Web site, http://www.nrc.gov 
(the Electronic Reading Room).

[[Page 71141]]

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

    Date of application for amendment: September 20, 2000.
    Brief description of amendment: The amendment allows placing a 
static VAR compensator into service with just one of the two protective 
subsystems operable.
    Date of issuance: November 13, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 136.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 2, 2000 (65 FR 
58829).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 13, 2000.
    No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of application for amendment: March 21, as supplemented on 
June 14, September 26, and October 16, 2000.
    Brief description of amendment: The proposed amendment revised the 
Technical Specifications to delete the reporting requirements for the 
core spray sparger inspection.
    Date of Issuance: November 2, 2000.
    Effective date: November 2, 2000 and shall be implemented within 30 
days of issuance.
    Amendment No.: 217.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: September 22, 2000 (65 
FR 57404).
    The June 14, September 26, and October 16, 2000, letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated November 2, 2000.
    No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of application for amendment: June 30, as supplemented on 
September 26, 2000.
    Brief description of amendment: The proposed amendment revised the 
Technical Specifications (TSs) to establish that the existing Safety 
Limit Minimum Critical Power Ratio contained in TS 2.1.A is applicable 
for the next operating cycle (Cycle 18).
    Date of Issuance: November 3, 2000.
    Effective date: November 3, 2000 and shall be implemented within 30 
days of issuance.
    Amendment No.: 218.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 22, 2000 (65 
FR 57406).
    The September 26, 2000, letter provided clarifying information 
within the scope of the original application and did not change the 
staff's initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated November 3, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 19, 1999, as 
supplemented July 18, 2000.
    Brief description of amendments: The amendments revise Technical 
Specification 5.5.11.c, ``Ventilation Filter Testing Program (VFTP),'' 
to include the requirement for laboratory testing of Engineered Safety 
Feature Ventilation System charcoal samples per American Society for 
Testing and Materials (ASTM) D3803-1989, ``Standard Test Method for 
Nuclear-Grade Activated Carbon,'' in response to Generic Letter 99-02, 
``Laboratory Testing of Nuclear-Grade Activated Charcoal,'' dated June 
3, 1999.
    Date of issuance: November 8, 2000.
    Effective date: November 8, 2000, and shall be implemented within 
45 days of the date of issuance.
    Amendment Nos.: Unit 1-130, Unit 2-130, Unit 3-130.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 8, 2000 (65 FR 
12287).
    The July 18, 2000, supplement provided clarifying information that 
was within the scope of the original application and Federal Register 
notice and did not change the staff's initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 8, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: February 29, 2000, as 
supplemented by letters dated June 26 and August 18, 2000.
    Brief description of amendments: The amendments revised the 
pressure-temperature (P-T) limits for heatup, cooldown, critical 
operation and inservice leak and hydrostatic test limitations for the 
reactor pressure vessel (RPV). The amendments replaced the current RPV 
P-T limit curves with three recalculated curves that are applicable to 
32 effective full power years. The staff has approved the revised 
limits for an interim period not to exceed December 15, 2002.
    Date of issuance: November 8, 2000.
    Effective date: Immediately until December 15, 2002, to be 
implemented within 30 days.
    Amendment Nos.: 144 and 130.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 5, 2000 (65 FR 
17911). The June 26 and August 18, 2000, submittals provided additional 
information that did not change the scope of the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 8, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 28, 2000.
    Brief description of amendments: The amendments revise License 
Condition 2.C.(37) for Unit 1 and License Condition 2.C.(21) for Unit 
2, to specify the types of fuel movements that cannot be performed 
during refueling unless all control rods are fully inserted.
    Date of issuance: November 9, 2000.
    Effective date: Immediately, to be implemented within 30 days.

[[Page 71142]]

    Amendment Nos.: 145 and 131.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Operating Licenses.
    Date of initial notice in Federal Register: June 14, 2000 (65 FR 
37422).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 9, 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: November 3, 1999, as 
supplemented by letters dated April 4, June 9, June 29, August 2, and 
August 16, 2000.
    Brief description of amendment: The amendment authorized revision 
of the Safety Analysis Report (SAR) to increase the containment 
structural design pressure from 54 psig to 59 psig, revised Technical 
Specification (TS) Table 3.3-3 to add a containment spray actuation 
signal on high-high containment building pressure to terminate main 
feedwater and main steam flow from the unaffected steam generator, 
revised TS 3.6.1.4 and Figure 3.6-1 to change the allowable containment 
initial conditions to be consistent with analysis assumptions, and 
revised TS 6.15 to increase the calculated peak accident pressure in 
the containment leakage rate testing program from 54 psig to 58 psig. 
Related changes to the Bases were also made.
    Date of issuance: November 13, 2000.
    Effective date: As of the date of issuance to be implemented prior 
to the commencement of heatup from refueling outage 2R14.
    Amendment No.: 225.
    Facility Operating License No. NPF-6: Amendment authorized revision 
to the SAR and revised the TSs.
    Date of initial notice in Federal Register: February 23, 2000 (65 
FR 9006).
    The June 29, 2000, supplement withdrew the proposed TS change to 
clarify the allowable containment leakage rate. The August 16, 2000, 
supplement withdrew the proposed TS change to increase the allowable 
containment spray pump degradation. The April 4, June 9, June 29, 
August 2, and August 16, 2000, supplemental letters provided clarifying 
information that was within the scope of the original Federal Register 
notice and did not change the staff's initial no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 13, 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: June 29, 2000, as supplemented 
by letter dated October 4, 2000.
    Brief description of amendment: The amendment revised the 
containment cooling system Technical Specifications to require that two 
independent containment cooling groups are operable with two 
operational cooling units in each group, in Modes 1, 2, 3, and 4.
    Date of issuance: November 13, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 226.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 26, 2000 (65 FR 
46008). The October 4, 2000, supplement provided clarifying information 
that was within the scope of the original Federal Register notice and 
did not change the staff's no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 13, 2000.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of application for amendments: January 19, 2000, as 
supplemented July 19, 2000.
    Brief description of amendments: The amendments will increase the 
setpoint tolerances for the pressurizer and main steam safety valves.
    Date of Issuance: November 14, 2000.
    Effective Date: November 14, 2000.
    Amendment Nos.: 166 and 110.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 5, 2000 (65 FR 
17915). The July 19, 2000, submittal provided clarifying information 
that did not change the scope of the original Federal Register Notice 
or change the initial no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 14, 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 26, 2000.
    Brief description of amendment: The amendment changes the Millstone 
Nuclear Power Station, Unit 3, Technical Specifications (TS) Section 
1.13, Definitions, ``Engineered Safety Features Response Time''; TS 
Section 1.28, ``Reactor Trip System Response Time''; TS Section 3.3.1, 
``Instrumentation-Reactor Trip System Instrumentation''; and TS Section 
3.3.2, ``Instrumentation-Engineered Safety Features Actuation System 
Instrumentation'' to provide for verification of response time for 
selected components provided that the components and the methodology 
for verification have been previously reviewed and approved by the 
Nuclear Regulatory Commission.
    Date of issuance: November 3, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 187.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 9, 2000 (65 FR 
48755).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 3, 2000.
    No significant hazards consideration comments received: No.

PECO Energy Company, PSEG Nuclear LLC, Delmarva Power and Light 
Company, and Atlantic City Electric Company

Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station Unit 
Nos. 2 and 3, York County, Pennsylvania

    Date of application for amendments: November 17, 1999, as 
supplemented June 15, 2000.
    Brief description of amendments: The amendments revise TS 5.5.7.c, 
``Ventilation Filter Testing Program (VFTP)'' to include the 
requirement for laboratory testing of Engineered Safety Feature 
Ventilation System charcoal samples per American Society for Testing 
and Materials D3803-1989 and the application of a safety factor of 2.0 
to the charcoal filter efficiency assumed in the plant design-basis 
dose analyses.
    Date of issuance: November 3, 2000.

[[Page 71143]]

    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendments Nos.: 237 and 240.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4288). The June 15, 2000, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination or expand the application beyond the scope 
of the original Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 3, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 12, 2000.
    Brief description of amendment: The proposed amendment would revise 
Technical Specification 3/4.7.5, ``Ultimate Heat Sink,'' by increasing 
the minimum required service water pond level from 415 feet to 416.5 
feet and decreasing the maximum allowed temperature at the discharge of 
the service water pumps from 95 degrees Fahrenheit to 90.5 degrees 
Fahrenheit.
    Date of issuance: November 14, 2000.
    Effective date: November 14, 2000.
    Amendment No.: 149.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 26, 2000 (65 FR 
46015).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 14, 2000.
    No significant hazards consideration comments received: No.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. Publicly available records will be 
accessible 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 December 29, 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, located at One White Flint North, 
11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly 
available records will be accessible 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

[[Page 71144]]

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, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, 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).

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

    Date of amendment request: November 2, 2000, as supplemented 
November 3, 2000.
    Description of amendment request: The amendment revises the 
Technical Specifications to allow reactor coolant system (RCS) 
inservice leak and hydrostatic testing to be performed with the reactor 
in the cold shutdown mode while the RCS temperature is greater than 212 
 deg.F (which normally corresponds to the hot shutdown mode).
    Date of issuance: November 3, 2000.
    Effective Date: As of its date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 267.
    Facility Operating License No. DPR-59: Amendment revises the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated 
November 3, 2000.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: Marsha Gamberoni.

    Dated at Rockville, Maryland, this 21st day of November 2000.
    For the Nuclear Regulatory Commission.
Suzanne C. Black,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 00-30282 Filed 11-28-00; 8:45 am]
BILLING CODE 7590-01-P