[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