Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations
Federal RegisterOct 20, 1999
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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
[[Page 56527]]
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 September 25, 1999, through October 7, 1999.
The last biweekly notice was published on October 6, 1999 (64 FR
54370).
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 Administration Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By November 19, 1999, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment
[[Page 56528]]
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: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendments request: September 14, 1999
Description of amendments request: Request No. 1: The proposed
administrative change to Technical Specification (TS) 5.5.2, Primary
Coolant Sources Outside Containment, would delete the references to the
post-accident sampling return piping of the radioactive waste gas
system and the post-accident sampling return piping of the liquid
radwaste system because the Palo Verde post-accident sampling system
does not have return lines to the radioactive waste gas or liquid
radwaste systems.
Request No. 2: This proposed TS amendment would also delete the
administrative requirement in TS 5.6.2, Annual Radiological
Environmental Operating Report, that states: ``[t]he report shall
identify the TLD [thermoluminescence dosimeter] results that represent
collocated dosimeters in relation to the NRC TLD program and the
exposure period associated with each result.''
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:
Request No. 1
Standard 1--Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No--This proposed administrative change to Technical
Specification (TS) 5.5.2 to delete references to the radioactive
waste gas system and liquid radwaste system in the context of the
post accident sampling system (PASS) does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. Leak testing requirements of the PASS return
piping are included in the TS 5.5.2 requirements that are not being
changed. The appropriate PASS piping, including return piping, is
leak tested per the prescribed requirements in TS 5.5.2. This
administrative change would simply clarify TS 5.5.2, since the PASS
return piping is not part of the waste gas or liquid radwaste
systems. There is no physical connection between the PASS piping and
the radioactive waste gas or liquid radwaste systems. The
radioactive waste gas system and the liquid radwaste system are not
part of PASS and would not contain highly radioactive fluids during
a serious transient or accident to be subject to TS 5.5.2. This
administrative change would involve no change to the design or
maintenance of the plant and no changes in the functional
requirements of any system.
Standard 2--Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No--This proposed administrative change to delete references to
the radioactive waste gas system and liquid radwaste system in the
context of PASS does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Leak testing requirements of the PASS return piping are implicitly
included in the TS 5.5.2 requirements that are not being changed.
The appropriate PASS piping, including return piping, is leak tested
per the prescribed requirements in TS 5.5.2. There is no physical
connection between the PASS piping and the radioactive waste gas or
liquid radwaste systems. The radioactive waste gas system and the
liquid radwaste system are not part of PASS and would not contain
highly radioactive fluids during a serious transient or accident to
be subject to TS 5.5.2. This administrative change would involve no
change to the design or maintenance of the plant and no changes in
the functional requirements of any system. This administrative
change would simply clarify TS 5.5.2, since the PASS return piping
is not part of the waste gas or liquid radwaste systems.
Standard 3--Does the proposed change involve a significant reduction in
a margin of safety?
No--This proposed administrative change does not involve a
significant reduction in a margin of safety. There is no margin of
safety associated with this proposed administrative change to
Technical Specification 5.5.2. Leak testing requirements of the PASS
return piping are implicitly included in the TS 5.5.2 requirements
that are not being changed. The appropriate PASS piping, including
return piping, is leak tested per the prescribed requirements in TS
5.5.2. This administrative change would involve no change to the
design or maintenance of the plant and no changes in the functional
requirements of any system. This administrative change would simply
clarify TS 5.5.2, since the PASS return piping is not part of the
waste gas or liquid radwaste systems.
Request No. 2
Standard 1--Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No--This proposed administrative change to Technical
Specification (TS) 5.6.2 does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
This proposed TS amendment would delete the administrative
requirement in TS 5.6.2, Annual Radiological Environmental Operating
Report, that states: ``[t]he report shall identify the TLD results
that represent collocated dosimeters in relation to the NRC TLD
program and the exposure period associated with each result.'' The
NRC ended their TLD program at the end of 1997. The requirements of
TS 5.6.2 and the changes being made with this request are purely
administrative reporting requirements that have no effect on the
design, operation, or maintenance of the plant. Since there is no
effect on the design, operation, or maintenance of the plant, this
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Standard 2--Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No--This proposed administrative change to TS 5.6.2 does not
create the possibility of a new or different kind of accident from
any accident previously evaluated. This change only affects
administrative reporting requirement and has no effect on the
design, operation, or maintenance of the plant. Since this proposed
change is purely administrative and would have no effect on the
design, operation, or maintenance of the plant, this change will not
create possibility of a new or different type of accident than any
previously evaluated.
[[Page 56529]]
Standard 3--Does the proposed change involve a significant reduction in
a margin of safety?
No--This proposed administrative change to TS 5.6.2 does not
involve a significant reduction in a margin of safety. This TS
establishes requirements for reporting radiological monitoring
information to the NRC. Since TS 5.6.2 contains an administrative
reporting requirement, and this proposed change would simply delete
an administrative requirement associated with a discontinued NRC
monitoring program, there is no margin of safety associated [with]
this TS or with the proposed changes to the requirements of TS
5.6.2. Also, since this involves only administrative reporting, this
change has no [e]ffect on any other margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Section Chief: Stephen Dembek
CBS Corporation (Licensee), Westinghouse Test Reactor, Waltz Mill Site,
Westmoreland, Pennsylvania, Docket No. 50-22, License No. TR-2
Date of amendment request: September 7, 1999, as supplemented on
October 1, 1999
Description of amendment request: CBS Corporation is the licensee
for the Westinghouse Test Reactor (WTR) at Waltz Mill, Pennsylvania.
The licensee is authorized to only possess the reactor and a
decommissioning plan has been approved. The licensee is planning to
revise the decommissioning plan by reassigning the responsibilities of
the Site Manager, who works for the Westinghouse Electric Company (a
contractor to CBS) to the TR-2 Decommissioning Project Director who
works for CBS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed amendment to a license of a facility involves no
significant hazards consideration if 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 the margin of safety.
The staff agrees with the licensee's no significant hazards
consideration determination submitted on September 7, 1999, for the
following reason:
In order to complete the decommissioning of the WTR facility as
described in the Decommissioning Plan, CBS has established contractual
agreements with the Westinghouse Electric Company to supply continued
site support and services to the Westinghouse Test Reactor Facility.
CBS has also entered into contracts with other third party
organizations as described in the Decommissioning Plan. These contracts
will remain in place between CBS and each respective third party so
that there will be no effective change in the personnel associated with
the on-going decommissioning project under the TR-2 License. CBS
continues to retain full responsibility for the project.
The only change being made is that the responsibilities of the
Westinghouse Electric Company Site Manager, as it pertains to the WTR
and the TR-2 License, has been assigned to the TR-2 Decommissioning
Project Director, who works for CBS. The Westinghouse Electric Company
personnel who reported to the Site Manager will now report directly to
CBS through the contract.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
The proposed amendment does not modify the WTR facility
configuration or licensed activities. Thus no new accident initiators
are introduced. Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated, and 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.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: September 16, 1999.
Description of amendment request: The amendments would revise
Surveillance Requirements (SRs) 3.8.4.8 and 3.8.4.9 of the Technical
Specifications and Bases SR 3.8.4.8 to allow testing of the direct
current (DC) channel batteries with the units on line. The proposed
change to SR 3.8.4.8 would also prohibit the diesel generator (DG)
batteries from being service tested while the units are on line.
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:
First Standard
Implementation of this amendment would not involve a significant
increase in the probability or consequences of an accident
previously evaluated. Approval of this amendment will have no
significant effect on accident probabilities or consequences. The
125 Volt DC Vital Instrumentation and Control Power System is not an
accident initiating system; therefore, there will be no impact on
any accident probabilities by the approval of this amendment. The
design of the system is not being modified by this proposed
amendment. It has been shown that the required battery testing can
be performed safely with the unit on line well within the allowed
outage time for an inoperable DC channel. Both safety trains would
continue to be capable of performing their required design functions
in the event of an accident. Therefore, there will be no impact on
any accident consequences.
Second Standard
Implementation of this amendment would not create the
possibility of a new or different kind of accident from any accident
previously evaluated. No new accident causal mechanisms are created
as a result of NRC approval of this amendment request. No changes
are being made to the plant which will introduce any new accident
causal mechanisms. This amendment request does not impact any plant
systems that are accident initiators.
Third Standard
Implementation of this amendment would not involve a significant
reduction in a margin of safety. Margin of safety is related to the
confidence in the ability of the fission product barriers to perform
their design functions during and following an accident situation.
These barriers include the fuel cladding, the reactor coolant
system, and the containment system. The performance of these fission
product barriers will not be impacted by implementation of this
proposed
[[Page 56530]]
amendment. It has already been shown that both safety trains of the
125 Volt DC Vital Instrumentation and Control Power System will
continue to be able to perform their accident mitigation functions
should they be required. In addition, the probabilistic risk
analysis conducted for this proposed amendment demonstrated that
there is no appreciable increase in overall plant risk incurred by
its implementation. No safety margins will be impacted.
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.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Attorney for licensee: Ms. Lisa F. Vaughn , Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina.
NRC Section Chief: Richard L. Emch, Jr.
Energy Northwest, Docket No. 50-397, WNP-2, Benton County, Washington
Date of amendment request: July 29, 1999, as supplemented by letter
dated August 30, 1999.
Description of amendment request: The proposed amendment would
delete a license condition that required installation of a neutron flux
monitoring system, in the form of excore wide range monitors (WRM), in
conformance with Regulatory Guide 1.97, ``Instrumentation for Light-
Water-Cooled Nuclear Power Plants to Assess Plant and Environs
Conditions During and Following an Accident.'' WNP-2 installed the WRM
system in the spring of 1989. Removal of the license condition would
allow WNP-2 to deactivate the WRM 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 an accident previously
evaluated.
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. As stated in the NRC safety evaluation approving NEDO-
31558-A (Reference 2) [in licensee's August 30,1999 letter],
Category 1 neutron flux monitoring instrumentation is not needed for
existing BWRs to cope with Loss-of-Coolant Accident (LOCA),
Anticipated Transient Without SCRAM (ATWS), or other accidents that
do not result in severe core damage conditions. Instrumentation to
monitor the progression of core melt accidents would best be
addressed by the current severe accident management program. Also,
WRM is not included in the WNP-2 IPE/PSA models and WRM is not
relied upon for operator actions in the Emergency Operating
Procedures (EOPs) or actions accounted for in Severe Accident
Management. Therefore, no individual precursors of an accident are
affected and the elimination of the WRM does not impact or change
the probabilities of accidents previously evaluated. In addition,
since the operability of plant systems designed to mitigate accident
consequence has not changed, the consequences of an accident
previously evaluated are not expected to increase.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors of
that accident. New accident precursors may be created by
modifications of the plant configuration, including changes in
procedures that may create the potential for new or different
personnel errors. The elimination of the WRM system does not create
the possibility of a new or different kind of accident because plant
crews are trained to use the Neutron Monitoring System (NMS) in
normal evolutions and under emergency conditions according to EOP
guidance. In addition, NEDO-31558-A concludes that the failure of
all neutron flux monitoring instrumentation does not prevent the
operator from determining the shutdown condition of the reactor.
Sufficient information is available on which to base operational
decisions and to conclude that reactivity control has been
accomplished. For example, Rod Position Information System (RPIS) is
powered from an uninterruptible source and remains available even
during Station Blackout (SBO) conditions to provide full core
control rod position information as a backup reactor power indicator
based on calculations of rod worth and shutdown margin. The proposed
change does not introduce any new modes of operation or alter system
setpoints which could create a new or different kind of accident.
Therefore, no new precursors of an accident and no new or different
kinds of accidents are created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The elimination of the WRM system does not result in a reduction
of the margin of safety. The neutron power indications necessary for
operator response to ATWS are provided by the NMS not WRM. Based on
a WNP-2 specific evaluation against the alternate criteria specified
in NEDO-31558-A, there is sufficient confidence that the
instrumentation would still be available to confirm that the reactor
is shutdown. In addition, failure of the existing neutron flux
monitoring instrumentation does not prevent plant operators from
determining the shutdown condition of the reactor. Sufficient
information is available to the operator to make operational
decisions and to conclude that reactivity control has been
accomplished. The proposed changes will not impact the basis for any
Technical Specification related to the establishment or maintenance
of nuclear safety margins. Therefore, operation of the facility in
accordance with the proposed amendment does not involve a reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502.
NRC Section Chief: Stephen Dembek.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: February 19, 1999.
Description of amendment request: The proposed amendment would
revise the Crystal River Unit 3 Improved Technical Specifications
Sections 5.6.2.7, 5.6.2.8, and 5.7.2.b, related to the Containment
Tendon Surveillance Program. The proposed changes are a result of
revisions to 10 CFR 50.55a which are required to be fully implemented
by September 9, 2001. These revised requirements affect the
surveillance methods for the containment tendons and the conduct of
containment visual inspections, and the methods of reporting the
results of the required inspections to the NRC.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
No. The proposed change to the Crystal River Unit 3 (CR-3)
Improved Technical Specifications (ITS) replaces the previous
programmatic commitment to implement a Containment Tendon
Surveillance Program based on Regulatory Guide 1.35, Revision 3,
[[Page 56531]]
with a Containment Inspection Program that complies with the current
requirements of 10 CFR 50.55a. Effective September 9, 1996, 10 CFR
50.55a requires licensees to implement a Containment Inspection
Program in compliance with the 1992 Edition with the 1992 Addenda of
Subsection IWE, ``Requirements for Class MC and Metallic Liners of
Class CC Components of Light-Water Cooled Power Plants,'' and with
Subsection IWL, ``Requirements for Class CC Concrete Components of
Light-Water Cooled Power Plants,'' of Section XI, Division 1, of the
American Society of Mechanical Engineers Boiler and Pressure Vessel
Code (ASME Code) with additional modifications and limitations as
stated in 10 CFR 50.55a(b)(2)(ix). Florida Power Corporation (FPC)
is implementing a Containment Inspection Program to comply with
these new regulatory requirements. The final rule specifies
requirements to assure that the critical areas of the containment
structure are routinely inspected to detect and take corrective
action for defects that could compromise structural integrity. This
proposed ITS change is requested to update the ITS to these latest
10 CFR 50.55a regulatory requirements.
By complying with the regulatory requirements described in 10
CFR 50.55a, the probability of a loss of containment structural
integrity is maintained as low as reasonably achievable. Maintaining
containment structural integrity is independent of the operation of
the reactor coolant system (RCS), and independent of the reactor
protection system (RPS) and emergency core cooling system (ECCS).
The Containment Inspection Program ensures that the containment will
function as designed to provide an acceptable barrier to release of
radioactive materials to the environment. By assuring the
effectiveness of this barrier through appropriate inspection, and by
implementing corrective actions for any degradation discovered
during these inspections that might lead to containment structural
failures, the probability or consequences of accidents will not be
greater than that previously evaluated.
2. Create the possibility of a new or different kind of accident
from previously evaluated accidents?
No. Maintaining containment structural integrity is independent
of the operation of the RCS, and independent of the RPS and ECCS. By
implementing corrective actions for any degradation discovered
during the required inspections of the containment, the possibility
of a new or different kind of accident will not be created.
3. Involve a significant reduction in a margin of safety?
No. The margin of safety as defined by the CR-3 ITS has not been
reduced. By complying with the regulatory requirements described in
10 CFR 50.55a, the probability of a loss of containment structural
integrity is maintained as low as reasonably achievable. The
Containment Inspection Program ensures that the containment will
function as designed to provide an acceptable barrier to release of
radioactive materials to the environment. By implementing the
Containment Inspection Program, the existing margin of safety is
preserved.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Attorney for licensee: R. Alexander Glenn, General Counsel (MAC-
BT15A), Florida Power Corporation, P. O. Box 14042, St. Petersburg,
Florida 33733-4042.
NRC Section Chief: Sheri R. Peterson.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of amendment request: July 7, 1999.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to change the component
surveillance frequencies for the following TSs to indicate a frequency
of once per 3 months: Core Spray System TS 4.4.A.1 and 4.4.A.2,
Containment Cooling System TS 4.4.C.1, Emergency Service Water System
TS 4.4.D.1, Fire Protection System TS 4.4.F (isolation valves only),
and Pressure Suppression Chamber--Drywell Vacuum Breakers TS 4.5.F.5.a.
The TSs currently stipulate a component surveillance frequency of once
per month. Also, the amendment would revise TS pages 4.4-1 and 4.4-2 to
incorporate editorial format changes and TS page 4.4-3 to accommodate
the expanded text.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed surveillance interval change does not alter the
actual surveillance requirements, nor does it alter the limits and
restrictions on plant operations. The reliability of systems and
components relied upon to prevent or mitigate the consequences of
accidents previously evaluated is not degraded by the proposed
change to the surveillance interval. Assurance of system and
equipment availability is maintained. The proposed change does not
alter any system or equipment configuration.
Based on the above, the proposed change does not significantly
increase the probability or consequences of an accident previously
evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed surveillance interval change does not alter the
actual surveillance requirements, nor does it alter the limits and
restrictions on plant operations. Assurance of system and equipment
availability is maintained. The proposed change does not alter any
system or equipment configuration nor does it introduce any new
mechanisms which could contribute to the creation of a new or
different kind of accident than previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed change extends the surveillance interval for
verifying the operability of the specified pumps and valves from
once per month to once per three months. The proposed change does
not alter the actual surveillance requirements, the limits and
restriction on plant operations nor the design, function or manner
of operation of any structures, systems or components. System
availability and reliability are maintained. Accordingly, the
proposed TS 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.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: S. Singh Bajwa.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: September 17, 1999.
Description of amendment requests: The proposed amendments would
allow credit in the applicable subcriticality analysis for the negative
reactivity provided by insertion of the rod cluster control assemblies
(RCCAs) during realignment from a cold leg recirculation to a hot leg
recirculation configuration. This realignment, which is referred to as
hot leg switchover, is performed following a loss-of-coolant accident.
This methodology change, when evaluated in accordance with 10 CFR
59.59, resulted in an unreviewed safety question that will require
prior approval by the NRC staff in accordance with the provisions of 10
CFR 50.90
[[Page 56532]]
prior to implementation. The proposed change would also affect the
Bases for Technical Specification (T/S) 3/4.5.5, ``Refueling Water
Storage Tank,'' and several sections of the Updated Final Safety
Analysis Report (UFSAR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated?
No. I&M [Indiana Michigan Power Company] proposes to credit RCCA
insertion of negative reactivity for criticality control during the
core cooling flow path realignment from cold leg recirculation to
hot leg recirculation following the postulated cold leg LBLOCA
[large-break loss-of-coolant accident]. No physical modifications
will be made to plant systems, structures, or components.
Credit for RCCAs is only being applied to demonstrate core
subcriticality upon hot leg switchover (HLSO) following a cold leg
LBLOCA. The performance criteria codified in 10 CFR 50.46 continue
to be met. The ability of the RCCAs to insert under LOCA and seismic
conditions was a function important to safety as part of the
original CNP [Cook Nuclear Plant] design basis. This is supported by
the conclusion presented in NRC (at the time, the Atomic Energy
Commission) Safety Evaluation Report (SER), Section 3.3,
``Mechanical Design of Reactor Internals,'' dated January 14, 1969.
The SER includes the statements that, ``[t]he control rod guide
tubes are designed so that each finger of each control rod assembly
is always partially inserted in the guide tube. Deflection limits on
the guide tubes have been chosen so that deflections caused by blow-
down forces during a loss-of-coolant accident will not prevent
control rod insertion,'' and that the ``* * * mechanical design of
internals, fuel assemblies, and control elements is acceptable.''
However, the licensing basis safety analyses for the LBLOCA scenario
have conservatively not taken credit for insertion of the RCCAs.
No physical modifications will be made to plant systems,
structures, or components in order to implement the proposed
methodology change. The safety functions of the safety related
systems and components, which are related to accident mitigation,
have not been altered. Therefore, the reliability of RCCA insertion
is not affected. As such, taking credit for RCCA insertion does not
alter the probability of an LBLOCA (the design basis accident at
issue). The Westinghouse analyses provided as Attachments 6 and 7
[to the licensee's application] demonstrate that RCCA insertion will
occur, with substantial margin, following a design basis cold leg
LBLOCA combined with a seismic event. Crediting RCCA insertion does
not affect mechanisms for a malfunction that could impact the HLSO
subcriticality analysis, or mechanisms that could initiate a LOCA.
Taking credit for the negative reactivity available from insertion
of the RCCAs, which is currently assumed for various accident
analyses within the CNP licensing basis (e.g., small break LOCA,
main steamline break, feedline break, steam generator tube rupture),
does not affect equipment malfunction probability directly or
indirectly. Therefore, crediting the RCCAs as a source of negative
reactivity for post-LOCA criticality control at the time of HLSO
does not significantly increase the probability of an accident
previously evaluated.
Furthermore, the traditional conservative assumption that the
most reactive RCCA is stuck fully out of the core is being
maintained. A malfunction that results in one RCCA to fail to insert
is a credible scenario, and is being considered for the post-LOCA
subcriticality analysis following a cold leg LBLOCA. There will be
sufficient negative reactivity, even with the most reactive RCCA
stuck fully out of the core, to assure core subcriticality post-
LOCA, as supported by the subcriticality analysis that is confirmed
each and every fuel cycle as part of the reload documentation (i.e.,
the Reload Safety Evaluations). The core is shown to remain
subcritical during the post-LOCA long-term cooling period,
specifically while HLSO is performed. Thus, no additional
radiological source terms are generated, and the consequences of an
accident previously evaluated in the UFSAR will not be significantly
increased.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. The proposed change involves crediting the negative
reactivity that is available from the RCCAs for an analysis
applicable several hours after the initiation of a cold leg LBLOCA.
As such, this change involves post-LOCA recovery actions several
hours after the break has occurred and does not involve accident
initiation. As discussed above, the original design requirements for
the CNP reactor internals, core fuel assemblies, and RCCAs were
based upon assuring the ability of the RCCAs to insert following a
double-ended rupture LOCA with seismic loadings. Thus, the safety
functions of safety related systems and components have not been
altered by this change. Crediting the negative reactivity that is
available from the RCCAs for the post-LOCA subcriticality analysis
upon HLSO does not cause the initiation of any accident, nor does
the proposed activity create any new credible limiting single
failure. Crediting the insertion of RCCAs does not result in any
event previously deemed incredible being made credible nor is there
any introduction of any new failure mechanisms that are not
currently considered in the design basis LOCA. There are no changes
introduced by this amendment concerning how safety related equipment
is designed to operate under normal or design basis accident
conditions since the calculations supporting RCCA insertion
following a cold leg LBLOCA have assumed design basis break sizes in
conjunction with seismic loadings. Therefore, the possibility of an
accident of a different type than already evaluated in the UFSAR is
not created.
3. Does the change involve a significant reduction in a margin
of safety?
No. Presently, no credit is taken for RCCA insertion in the
analysis to demonstrate post-cold leg LOCA subcriticality at the
time of HLSO. The current subcriticality analysis for this scenario
relies only on the boron provided by the RWST [refueling water
storage tank] and the accumulators. Thus, RCCA insertion provides
another source of negative reactivity (margin of safety). Revising
the post-cold leg LBLOCA HLSO subcriticality analysis to credit the
negative reactivity associated with the RCCAs is a means to offset
the sump dilution associated with the effects of the inactive
regions of the CNP containment sump. The incorporation of this
``defense-in-depth'' source of negative reactivity in the HLSO
subcriticality analysis has been conservatively determined to cause
a reduction in the margin of safety. 10 CFR 50, Appendix K, I.A.2.,
states, in part, that ``[r]od trip and insertion may be assumed if
they are calculated to occur,'' and provides for crediting RCCA
insertion as an acceptable feature of emergency core cooling system
(ECCS) evaluation models. The proposed change is based upon an
analysis for CNP that demonstrates that the control rods will indeed
insert and the resulting negative reactivity can be credited for
post-LOCA criticality control.
The proposed change would ensure that post-LOCA subcriticality
is maintained during HLSO. Subsequently, there would not be a
challenge to long-term core cooling due to a return to a critical
condition. This being the case, the requirements of 10 CFR
50.46(b)(5) that, ``* * * the calculated core temperature shall be
maintained at an acceptably low value and decay heat shall be
removed for the extended period of time* * *'' continues to be
satisfied and the margin of safety in the CNP licensing basis is
preserved. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085.
Attorney for licensee: Jeremy J. Euto, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: Claudia M. Craig.
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: September 29, 1999.
Description of amendment request: The proposed amendment requests a
Technical Specification change that
[[Page 56533]]
would extend the allowed out-of-service time for the residual heat
removal service water system (RHRSW) from 7 days to 11 days on a one-
time basis while modifications are made on the RHRSW ``A'' strainer.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92 since it would not:
Involve an increase in the probability or consequences of an
accident previously evaluated.
The Conditional Core Damage Probability due to this proposed
change is calculated to be 6.4 E-8. This value falls below the
threshold probability of 1 E-6 for risk significance of temporary
changes to the plant configuration in the EPRI PSA [Electric Power
Research Institute Probability Assessment] Applications Guide
(Reference 3) [see application dated September 29, 1999].
This proposed change does not increase the consequences of an
accident previously evaluated because all relevant accidents (LOCA)
[loss-of-coolant accident] would result in the transfer of decay
heat to the suppression pool. For this scenario, the same complement
of equipment will be available to achieve and maintain cold shutdown
as is required by the current Technical Specification LCO [limiting
condition for operation].
Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change does not physically alter the plant. As
such, no new or different types of equipment will be installed. The
new design for the RHRSW strainer packing gland will be evaluated
under a separate 10 CFR 50.59 evaluation and is considered to be
functionally equivalent for the purposes of this one-time-only
proposed Technical Specification change.
The implementation and use of the contingency plan for achieving
limited containment heat removal in the event the B division of
RHRSW is rendered inoperable will be evaluated under the Authority's
10 CFR 50.59 program.
Involve a significant reduction in a margin of safety.
The Conditional Core Damage Probability due to this proposed
change is calculated to be 6.4 E-8. This value falls below the
threshold probability of 1 E-6 for risk significance of temporary
changes to the plant configuration in the EPRI PSA Applications
Guide (Reference 3).
The consequences of a postulated accident occurring during the
extended allowable out-service time are bounded by existing analyses
therefore there is 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Section Chief: S. Singh Bajwa.
Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364,
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama
Date of amendment request: December 1, 1998, as supplemented by
letters of April 21, 1999, and July 19, 1999.
Description of amendment request: The proposed amendments would
revise the Technical Specifications to reflect replacing the current
Model 51 steam generators with Westinghouse Model 54F steam generators.
The replacement program includes re-analyzing and evaluating loss-of-
coolant-accident (LOCA) and non-LOCA mass and energy releases,
containment and sub-compartment pressure and temperature responses,
dose analyses, and the effects on nuclear steam supply and balance of
plant systems.
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 changes do not significantly increase the
probability or consequences of an accident previously evaluated in
the [Final Safety Analysis Report] FSAR. The comprehensive
engineering effort performed to support [steam generator] SG
replacement has included evaluations or re-analysis of all accident
analyses including all dose related events. All dose consequences
have been analyzed or evaluated with respect to these proposed
changes, and all acceptance criteria continue to be met. Therefore,
these changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident than any accident already evaluated in
the FSAR. No new accident scenarios, failure mechanisms or limiting
single failures are introduced as a result of the proposed changes.
The proposed technical specification changes have no adverse effects
on any safety-related system and do not challenge the performance or
integrity of any safety-related system. Therefore, these changes do
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed technical specification changes do not involve a
significant reduction in a margin of safety. All applicable analyses
supporting the [steam generator] SG replacement reflect these
proposed values. All acceptance criteria (including LOCA peak clad
temperature, [departure from nucleate boiling] DNB, containment
temperature and pressure, and dose limits) continue to be met.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed Southern Nuclear Company'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.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama.
NRC Section Chief: Richard L. Emch, Jr.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, (SQN), Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: June 30, 1999 (TS 98-10).
Description of amendment requests: The proposed amendments would
change the Sequoyah (SQN) Operating Licenses DPR-77 (Unit 1) and DPR-
79(Unit 2) by updating the current Technical Specification requirements
for reactor coolant system leakage detection and operational leakage
specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), 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 revisions enhance the Technical specification (TS)
requirements to provide greater consistency with the standard TS in
NUREG-1431. This revision proposes changes to the requirements for
reactor coolant system (RCS) leak detection and RCS operational
leakage in Specifications 3.4.6.1 and 3.4.6.2, respectively. New
Specifications
[[Page 56534]]
3.4.6.3 and 3.5.6 for RCS pressure isolation valves and emergency
core cooling system (ECCS) seal injection flow have been added to
improve consistency with NUREG-1431. The proposed revisions are not
the result of changes to plant equipment, system design, testing
methods, or operating practices. The modified requirements will
allow some relaxation of current operability criteria, action
requirements, and surveillance requirements (SRs). These changes
provide more appropriate requirements in consideration of the safety
significance and the design capabilities of the plant as determined
by the improved standard TS industry effort. These specifications
serve to primarily provide identification and control of the RCS
fission product barrier leakage and ECCS degradation and are not
considered to be a contributor to the generation of postulated
accidents. Since these proposed revisions will continue to support
the required safety functions, without modification of the plant
features, the probability of an accident is not increased.
The proposed changes will allow relaxation of action times for
inoperable leak detection features and the components that can be
inoperable. The required actions to ensure acceptable pressure
isolation valve capability with an inoperable valve have been
revised to allow isolation by a single valve for a limited period of
time. These revisions will allow unit operation for a longer period
of time with reduced system redundancy. However, the redundancy
reduction and action time increases are not significant and will
continue to provide an acceptable level of safety considering the
significance of RCS leakage, other design features or compensatory
actions that provide equivalent functions, and the unlikely chance
of an event that would require functions for leakage identification
during the proposed time interval. These considerations are
consistent with the basis developed by the industry and NRC for
NUREG-1431. Surveillances have been removed from the RCS operational
leakage specification as a result of relocated requirements,
duplication of other SRs, and testing requirements that do not
provide a significant benefit in the identification of RCS leakage.
The SRs that have been retained or relocated to other TS
specifications will provide acceptable verifications for the timely
identification of conditions that indicate an unacceptable amount of
RCS leakage or potential ECCS degradation resulting from excessive
seal injection flow.
The limiting condition for operation associated with the seal
injection flow requirements has been revised to utilize a modified
operability criteria. The proposed change will provide a range of
differential pressures and the corresponding seal flows that would
be representative of the existing single point flow limit. This
change does not alter the intent of the operability requirements,
but does allow the flexibility to use equivalent values that provide
the same level of assurance for ECCS operability. The proposed
operability condition for seal injection flow enhances the current
requirement by establishing additional test parameters that will
ensure that the amount of seal injection flow does not degrade the
ECCS functions.
The proposed changes to the SQN TS provide flexibility without
modifying the functions of required safety systems. In many
instances the proposed changes ensure that plant conditions for
surveillance testing are more appropriate for testing purposes and
the verification of system operability.
These changes are consistent with the intent of NUREG-1431 and
result in the enhancement of the SQN TSs based on the latest
industry and NRC positions. The provisions proposed in this change
request will continue to maintain an acceptable level of protection
for the health and safety of the public and will not significantly
impact the potential for the offsite release of radioactive
products. The overall effect of the proposed change will result in
specifications that have equivalent or improved requirements
compared to existing specifications for RCS leakage and ECCS
operability and will not significantly increase the consequences of
an accident.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed revisions are not the result of changes to plant
equipment, system design, testing methods, or operating practices.
The modified requirements will allow some relaxation of current
operability criteria, action requirements, and SRs consistent with
NUREG-1431. These changes provide more appropriate requirements in
consideration of the safety significance and the design capabilities
of the plant as determined by the improved standard TS industry
effort. These specifications serve to primarily provide
identification and control of the RCS fission product barrier
leakage and ECCS degradation and are not considered to be a
contributor to the generation of postulated accidents. Since the
functions of the associated systems will continue to perform without
change and were not previously considered to contribute to accident
generation, the proposed changes will not create the possibility of
a new or different kind of accident.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed changes, associated with RCS leakage and ECCS
functions, will not result in changes to system design or setpoints
that are intended to ensure timely identification of plant
conditions that could be precursors to accidents or potential
degradation of accident mitigation systems. These systems will
continue to operate without change and only the associated actions
or testing activities have been altered. Revisions to the actions
and surveillances provide some relaxation and flexibility such that
longer intervals are allowed for inoperable components and testing
requirements are revised to provide conditions that provide more
accurate results. The increased action times are acceptable
considering the available redundant features, the compensatory
measures provided by the actions, and the allowed time intervals
that have been developed by the industry and NRC and recommended in
NUREG-1431. The SR changes actually provide test condition
requirements that enhance the accuracy of the activity even though
they may allow a delay in the performance of the test. These
surveillance changes are also in accordance with NUREG-1431
recommendations.
These revisions will continue to provide the necessary actions
to minimize the impact of inoperable equipment to an acceptable
level and will provide testing activities that will ensure system
operability. Since the setpoints and design features that support
the margin of safety are unchanged and actions for inoperable
systems continue to provide appropriate time limits and compensatory
measures, the proposed changes will not significantly reduce the
margin of safety.
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Sheri R. Peterson.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: September 28, 1999 (TS 99-007).
Description of amendment request: The proposed amendment on
Response Time Test (RTT) elimination would revise the Watts Bar Nuclear
Plant Unit 1 Technical Specifications (TS) definitions for ``Engineered
Safety Feature (ESF) Response Time'' and ``Reactor Trip System (RTS)
Response Time'' 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 NRC.
In addition, associated changes to the Bases for Surveillance
Requirements would also be made.
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.
This change to the TS does not result in a condition where the
design, material, and
[[Page 56535]]
construction standards that were applicable prior to the change are
altered. The same RTS and ESF instrumentation is being used, the
time response allocations/modeling assumptions in the Chapter 15
analyses are unchanged; only the method of verifying time response
is changed. The proposed change will not modify any system interface
and could not increase the likelihood of an accident since these
events are independent of this change. The proposed activity will
not change, degrade or prevent actions, or alter any assumptions
previously made in evaluating the radiological consequences of an
accident described in the UFSAR [Updated Final Safety Analysis
Report]. 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.
This change does not alter the performance of pressure and
differential pressure transmitters, process protection racks (Eagle
21), nuclear instrumentation (NIS), and logic system (SSPS) used in
the plant protection systems. These components/systems will still
have response time verified by test prior to placing the equipment
in operational service and after any maintenance that could affect
the response time of that equipment. Changing the method of
periodically verifying instrument response time for applicable
instrumentation from RTT to calibration and channel checks or
functional test will not create any new accident initiators or
scenarios. Therefore, the proposed amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
This change does not affect the total system response time
assumed in the safety analysis. The periodic system response time
verification method for selected pressure and pressure differential
sensors, Eagle 21, NIS, and SSPS is modified to allow use of actual
test data or engineering data. The method of verification still
provides assurance that the total system response time is within
that assumed in the safety analysis, since calibration checks and
functional tests will detect any degradation which might
significantly affect equipment response time. Therefore, the
proposed license amendment request does not result in a significant
reduction in 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.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
NRC Section Chief: Sheri Peterson.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: June 15, 1999.
Description of amendment request: The licensee proposed revisions
to Technical Specifications (TSs) Sections 3.1/4.1 Reactor Protection
System and 3.2/4.2 Protective Instrument Systems instrumentation,
tables, and the associated bases to increase the surveillance test
intervals (STIs), add allowable out-of-service times (AOTs), replace
generic ECCS actions for inoperable instrument channels with function-
specific actions, and relocate selected trip functions from the TSs to
a Vermont Yankee (VY) controlled document. In addition, revision to TS
Section 3.1/4.1 Reactor Protection System and the associated bases is
proposed to remove the RUN Mode APRM Downscale/IRM High Flux/
Inoperative Scram Trip Function (APRM Downscale RUN Mode SCRAM). The
submittal also proposes to implement editorial corrections and
administrative changes that do not alter the meaning or intent of the
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 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.
VY has determined that the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated. The generic analysis contained in
Licensing Topical Report NEDC-30851P-A assessed the impact of
changing SCRAM (RPS) surveillance test intervals for Logic and
Functional tests (STIs) and adding allowable out-of-service times
(AOTs) on the SCRAM (RPS) failure frequency, the scram frequency and
equipment cycling. Specifically, Section 5.7.4, ``Significant
Hazards Assessment,'' of NEDC-30851P-A states that:
``Fewer challenges to the safeguards system, due to less
frequent testing of the RPS, conservatively results in a decrease of
approximately one percent in core damage frequency. This decrease is
based upon the following:
Based on the plant-specific experience presented in Appendix J,
the estimated reduction in scram frequency (0.3 scrams/ yr.)
represents a 1 to 2 percent decrease in core damage frequency based
on the BWR plant-specific Probabilistic Risk Assessments (PRAS)
listed in Table 5-8.
The increase in core damage frequency due to less frequent
testing is less than one percent. This increase is even lower (less
than 0.01 percent) when the changes resulting from the
implementation of the Anticipated Transients Without Scram (ATWS)
rule are considered. Therefore, this increase is more than offset by
the decrease in CDF due to fewer scrams.
The effect of reducing unnecessary cycles on RPS equipment,
although not easily quantifiable, also results in a decrease in core
damage frequency.
The overall impact on core damage frequency of the changes in
allowable out-of-service times is negligible.''
From this generic analysis, the BWR Owners' Group concluded that
the proposed changes do not significantly increase the probability
or consequences of an accident previously evaluated, namely the
increase in probability of a scram failure due to SCRAM (RPS)
unavailability is insignificant, and the overall probability of an
accident is actually decreased as the time the SCRAM (RPS)
Instrumentation logic operates as designed is increased resulting in
less inadvertent scrams during testing and repair. Furthermore, the
plant specific reports demonstrate[ ] that although VY differs from
the generic model analyzed in License Topical Report NEDC-30851P-A,
the net effect of the plant-specific differences do not alter the
generic conclusions.
The generic analysis contained in Licensing Topical Reports
NEDC-30851P-A Suppl 2/NEDC-31677P-A assessed the impact of changing
STIs and AOTs for BWR Isolation Instrumentation common/not common to
SCRAM (RPS) and ECCS instrumentation. Specifically, Section 4.0,
``Summary of Results,'' of NEDC-30851P-A Suppl 2 states that:
``The results indicate that the effects on probability of
failure to initiate isolation are very small and the effects on
probability or frequency of failure to isolate are negligible in
nearly every case. In addition, the results indicate that increasing
the AOT to 24 hours for tests and repairs has a negligible effect on
the probability of failure of the isolation function. These combined
with changes to the testing intervals and allowed out-of-service
times for RPS and ECCS instrumentation provide a net improvement to
plant safety and operations.''
and Section 5.6, ``Assessment of Net Effect of Changes,'' of NEDC-
31677P-A states that:
``A reduction in core damage frequency (CDF) of at least as much
as estimated in the ECCS instrumentation analysis can be expected
when the isolation actuation instrumentation STIs are changed from
one month to three months. The chief contributor to this reduction
is the channel functional tests for the MSIVs. Inadvertent closure
of the MSIVs will cause an unnecessary plant scram. This reduction
in CDF more than compensates for any small incremental
[[Page 56536]]
increase (10% or 1OE-07/year) in calculated isolation function
failure frequency when the STI is extended to three months.''
From this generic analysis, the BWR Owners' Group concluded that
the proposed changes do not significantly increase the consequences
of an accident previously evaluated, namely the increase in
probability of an isolation failure due to isolation instrumentation
unavailability is insignificant, and the overall probability of an
accident is actually decreased as the time the SCRAM (RPS)
Instrumentation logic operates as designed is increased resulting in
less inadvertent scrams during testing and repair.
The generic analysis contained in Licensing Topical Report NEDC-
30936P-A (Parts 1 and 2) assessed the impact of changing STIs and
AOTs for all BWR ECCS Actuation Instrumentation. Specifically,
Section 4.0, ``Technical Assessment of Changes,'' of NEDC-30936P-A
(Part 2) states that:
``The results indicate an insignificant (less than 5E-7 per
year) increase in water injection function failure frequency when
STIs are increased from 31 days to 92 days, AOTs for repair of the
ECCS actuation instrumentation are increased from one hour to 24
hours, and AOTs for surveillance testing are increased from two to
six hours. For all four BWR models the increase represents less than
4% increase in failure frequency. However, when other factors which
influence the overall plant safety are considered, the net result is
judged to be an improvement in plant safety.''
From this generic analysis, the BWR Owners' Group concluded that
the proposed changes do not significantly increase the probability
or consequences of an accident previously evaluated, namely the
increase in probability of a water injection failure due to ECCS
instrumentation unavailability is insignificant and the net result
is judged to be an improvement in plant safety. Furthermore, the
plant specific report demonstrates that although VY differs from the
generic model analyzed in Licensing Topical Report NEDC30936P-A, the
net affect of the plant-specific differences do not alter the
generic conclusions.
The generic analysis contained in Licensing Topical Report NEDC-
30851 P-A Supp 1, assessed the impact of changing Rod Block STIs on
Rod Block failure frequency. Specifically, Section 5 (BNL's Tech.
Eval. Report--Attach. 2 to the NRC SER) of NEDC-30851 P-A Suppl 1
states that:
``The BWR Owners'' Group proposed changes to the Technical
Specifications concerning the test requirements for BWR control rod
block instrumentation. The changes consist of increasing the
surveillance test intervals from one to three months. These test
interval extensions are consistent with the already approved changes
to STIs for the reactor protection system. The technical analysis
reviewed and verified as documented herein indicates that there will
be no significant changes in the availability of the control rod
block function if these changes are implemented. In addition, there
will be a negligible impact on the plant core melt frequency due to
the decreased testing.''
From this generic analysis, the BWR Owners' Group concluded that
the proposed changes do not significantly increase the probability
of an accident previously evaluated or consequences of an accident
previously evaluated.
Bases contained in GE Topical Report GENE-770-06-1 assessed the
impact of changing STIs and AOTs on selected systems failure
frequency. Specifically, Section 2.0, ``Summary,'' of GENE 770-06-1
states that:
``Technical bases are provided for selected proposed changes to
the instrumentation STIs and AOTs that were identified in the BWROG
Improved BWR Technical Specification activity. These STI and AOT
changes are consistent with approved changes to the RPS, ECCS, and
isolation actuation instrumentation. These proposed changes do not
result in a degradation to overall plant safety.''
From these Bases, the BWR Owners' Group concluded that the
proposed changes do not significantly increase the probability of an
accident previously evaluated or consequences of an accident
previously evaluated.
Bases contained in GE Topical Report GENE-770-06-2 assessed the
impact of changing STIs and AOTs on selected systems (RCIC
Actuation) failure frequency. Specifically, Section 2.0,
``Summary,'' of GENE 770-06-2 states that:
``The STI and AOT changes to the RCIC actuation instrumentation
are justified based on their small effect on the water injection
function unavailability and consistency with comparable changes to
the actuation instrumentation for the other ECCS subsystems''. These
STI and AOT changes are consistent with approved changes to the RPS,
ECCS, and isolation actuation instrumentation. These proposed
changes do not result in a degradation to overall plant safety.''
From these Bases, the BWR Owners' Group concluded that the
proposed changes do not significantly increase the probability of an
accident previously evaluated or consequences of an accident
previously evaluated.
The proposed change will not alter the physical characteristics
of any plant systems or components and all safety-related systems
and components remain within their applicable design limits. Thus,
system and component performance is not adversely affected by this
change, thereby assuring that the design capabilities of those
systems and components are not challenged in a manner not previously
assessed so as to create the possibility of a new or different kind
of accident.
The addition of allowable out-of-service times (AOTs) and the
increase in surveillance test intervals (STIS) does not alter the
function of the SCRAM (RPS), ECCS, Isolation, Rod Block, and
Selected Instrument Systems nor involve any type of plant
modification and no new modes of plant operation are involved with
these changes.
No physical change is being made to any systems or components
that are credited in the safety analysis, therefore there is no
change in the probability or consequences of any accident analyzed
in the UFSAR.
The design basis accident applicable to the startup power region
is the Control Rod Drop Accident (CRDA). The UFSAR does not credit
the RUN Mode IRM High Flux/Inoperative with the associated APRM
downscale scram Trip Function (APRM downscale RUN Mode SCRAM) in the
termination of this accident, Accident mitigation is provided by the
APRM 120% power scram. Therefore, elimination of the APRM downscale
RUN Mode SCRAM function has no adverse affect on previously
evaluated accidents.
The Continuous Control Rod Withdrawal Error (CWE) transient is
terminated by the Rod Block Monitor (RBM) in the RUN Mode. The APRM
Reduced High Flux Scram provides the primary STARTUP Mode protection
in conjunction with the IRMs and limits the consequences of this
transient. Therefore, elimination of the APRM downscale RUN Mode
SCRAM function has no effect on the consequences of this transient.
Adding a new surveillance to verify SRM/IRM/APRM will enhance
neutron monitoring during startups and shutdowns and does not have
an adverse affect on previously evaluated accidents.
None of the proposed changes will affect any of the rod blocks
or other precursor events to either the CRDA or CWE. Therefore,
there is no change in the probability of any accident previously
analyzed.
Use of ECCS Function-specific AOTs, actions and relocation of
Bus Power Monitors to a licensee controlled document is consistent
with STS and does not have an adverse affect on previously evaluated
accidents.
In addition, VY concluded the editorial corrections and
administrative changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
These changes do not alter the meaning or intent of any
requirements.
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 an accident
previously evaluated.
VY has determined that the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change will not alter the physical characteristics
of any plant systems or components and all safety-related systems
and components remain within their applicable design limits. Thus,
system and component performance is not adversely affected by this
change, thereby assuring that the design capabilities of those
systems and components are not challenged in a manner not previously
assessed so as to create the possibility of a new or different kind
of accident. Editorial corrections and administrative changes do not
alter the meaning or intent of any requirements.
The addition of allowable out-of-service times (AOTs), ECCS
function-specific actions and the increase in surveillance test
intervals (STIs) does not alter the function of the SCRAM (RPS),
ECCS, Isolation, Rod Block,
[[Page 56537]]
and Selected Instrument Systems nor involve any type of plan
modification and no new modes of plant operation are involved with
these changes. Therefore, operation in accordance with the proposed
amendment will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
Elimination of APRM downscale RUN Mode SCRAM function affects
only the operations of neutron monitoring and protective systems
(IRM and APRM) which provide indication and mitigation actions only.
Operation of these systems does not create the possibility for new
precursors (such as reactivity) which would introduce a new or
different kind of accident from any accident previously evaluated.
Additionally, the proposed changes do not affect the ability of
those systems required to mitigate previously evaluated accidents
during the modes they are credited.
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 The NRC staff has
reviewed and approved the generic studies contained in the GE
Topical Reports (LTRs) and has concurred with the BWR Owners' Group
that the proposed changes do not significantly affect the
availability of the SCRAM (RPS), ECCS, Isolation, Rod Block, or
Selected Instrument Systems. The proposed addition of allowable out-
of-service times (AOTs) for the instruments addressed in the LTRs
provide reasonable time for making repairs and performing tests. The
lack of sufficient AOTs in the current Technical Specifications (TS)
creates a hurried atmosphere during repairs and tests that could
cause an increased risk of error. In addition, placing an individual
channel in a tripped condition because no AOT exists, as in the
current TS, increases the potential of an inadvertent scram. The
proposed AOTs provide realistic times to complete the required
actions without increasing the overall instrument failure frequency.
Use of ECCS Function-specific AOTs, actions and relocation of Bus
Power Monitors to a licensee controlled document is consistent with
STS and there is no significant reduction in the margin of safety.
Editorial corrections and administrative changes do not alter
the meaning or intent of any requirements. Therefore, there is no
significant reduction in the margin of safety.
The incorporation of extended surveillance test intervals (STIs)
does not result in significant changes in the probability of
instrument failure, as demonstrated by the LTRs. In addition, the TS
calibration frequency has not changed, and therefore assurance
exists that the setpoints will not be affected by drift.
These changes, when coupled with the reduced probability of
test-induced plant transients and equipment failures, result in an
overall increase in the margin of safety.
The only scram function that the UFSAR takes credit for in the
mitigation of the limiting accident (control rod drop accident) is
the APRM 120% power scram which is not affected by this change. Only
the APRM Downscale RUN Mode SCRAM, for which the UFSAR takes no
credit in the termination of any analyzed event, is removed by this
change. Removal of the APRM Downscale RUN Mode SCRAM will avoid the
need to operate the plant in a ``half scram'' condition with the
potential for an inadvertent plant transient. For these reasons, the
change does not involve a significant reduction in a margin of
safety.
The Continuous Control Rod Withdrawal Error (CWE) transient is
terminated by the Rod Block Monitor (RBM) in the RUN Mode. When
initiated from the STARTUP Mode, the consequences of a CWE are
limited by the APRM Reduced High Flux scram in conjunction with the
IRM scram function. Therefore eliminating the TS requirement for the
APRM Downscale RUN Mode SCRAM will not reduce the margin of safety
for this transient.
Adding a new surveillance to verify SRM/IRM/APRM overlap will
enhance neutron monitoring during startups and shutdown, and
consequently does not involve a significant reduction in a margin of
safety.
On the basis of the above, VY has determined that operation of
the facility in accordance with the proposed change does not involve
a significant hazards consideration as defined in 10 CFR 50.92(c),
in that it: (1) does not involve a significant increase in the
probability or consequences of an accident previously evaluated; (2)
does not create the possibility of a new or different kind of
accident from any accident previously evaluated; and (3) 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.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301
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.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: September 21, 1999.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) 3.10.C, ``Diesel Fuel'' by
increasing the minimum usable volume of diesel fuel in the diesel fuel
oil storage tank (FOST). The specified minimum amount of diesel fuel is
that quantity necessary to support diesel generator operation for a
period of 7 days.
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. Will the proposed changes involve a significant increase in
the probability or consequences of an accident previously evaluated?
The diesel generators are used to support mitigation of the
consequences of an accident; however, they are not considered the
initiator of any previously analyzed accident. This change does not
challenge or degrade the performance of any safety system assumed to
function in the accident analysis. Since this change simply
increases the minimum volume of stored diesel generator fuel in the
FOST, its impact is to enhance the long-term operation of diesel
generators used to mitigate the consequences of accidents.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Will the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
This change does not affect the design or mode of operation of
any plant system, structure or component. No physical alteration of
plant structures, systems or components is involved, and no new or
different type of equipment will be installed. Thus, no new
condition of operation is created. The change is conservative in
that it results in a net increase in the minimum required diesel
fuel oil stored in the FOST.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated for Vermont Yankee.
3. Will the proposed changes involve a significant reduction in
a margin of safety?
The[ ] proposed change does not adversely affect a margin of
safety because increasing the minimum required volume of fuel oil
provides additional assurance of diesel generator availability and,
therefore, maintains or increases the availability of the onsite
power supply. Since this change simply increases the quantity of
diesel fuel oil available for diesel generator operation, there is
no reduction in any value, condition, or range of parameters used in
any accident analysis.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
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.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
[[Page 56538]]
NRC Section Chief: James W. Clifford.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: September 21, 1999.
Description of amendment request: The proposed amendment would
extend the effective full implementation date by six months, from
December 31, 1999, to June 30, 2000, for Amendment 120 issued March 22,
1999. Amendment 120 approved a modification to the plant to increase
the storage capacity of the spent fuel pool and increase the nominal
fuel enrichment to 5 weight percent U-235. The extension is due to
delays fabricating and installing the new spent fuel storage racks.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change is administrative in nature and does not
significantly affect any system that is a contributor to initiating
events for previously evaluated accidents. The proposed change does
not significantly affect any system that is used to mitigate any
previously evaluated accidents. Therefore, the proposed change does
not involve any significant increase in the probability or
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.
The proposed change is administrative in nature and does not
alter the design, function, or operation of any plant component and
does not install any new or different equipment. Therefore, a
possibility of a new or different kind of accident from those
previously analyzed has not been created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change is administrative in nature and does not
involve a significant reduction in the margin of safety associated
with the fuel cladding, reactor coolant boundary, containment, or
any safety limit.
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.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Section Chief: Stephen Dembek.
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.
Consolidated Edison Company of New York, Docket No. 50-003, Indian
Point Nuclear Generating Station, Unit No. 1, Buchanan, New York
Date of amendment request: July 20, 1999.
Description of amendment request: The amendment would revise the
Technical Specifications to change the senior reactor license
requirement for the Operations Manager.
Date of publication of individual notice in Federal Register:
September 9, 1999 (64 FR 49027).
Expiration date of individual notice: October 12, 1999.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: September 24, 1999.
Description of amendment request: The proposed amendment would
revise current Technical Specification (TS) 3.6.1.8 by adding footnote
``**'' to Action b. The footnote would allow continued operation of
Fermi 2 with the leakage of penetration X-26 exceeding the limit in TS
4.6.1.8.2, provided certain compensatory measures are taken. Operation
would be allowed to continue until the next plant shutdown.
Because the NRC staff issued the Fermi 2 improved standard TSs
(ITS) on September 30, 1999, with implementation within 90 days, the
licensee also provided a version of the TS amendment that would be
compatible with the ITS. This version would add a new special
operations TS, ITS 3.10.8, to address the compensatory actions and
other requirements associated with penetration X-26.
Date of publication of individual notice in Federal Register:
October 1, 1999 (64 FR 53421).
Expiration date of individual notice: Comment period expires
October 15, 1999; Opportunity for hearing period expires November 1,
1999.
Local Public Document Room location: Monroe County Library System,
Ellis Reference and Information Center, 3700 South Custer Road, Monroe,
Michigan 48161.
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
[[Page 56539]]
Evaluation and/or Environmental Assessment as indicated. All of these
items are available for public inspection at the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
and at the local public document rooms for the particular facilities
involved.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: May 20, 1999, as supplemented
by letters dated September 8, 1999, September 16, 1999, and September
20, 1999.
Brief description of amendments: The amendments revised Technical
Specification (TS) Section 3.8.A, ``Containment Cooling Service Water
System,'' (CCSW) to clarify that only one pump is required to support
operability of the Control Room Emergency Ventilation System (CREVS).
Date of issuance: October 1, 1999.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 174 and 170.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 25, 1999 (64 FR
46426). The September 8, September 16, and September 20, 1999,
submittals provided additional clarifying information that did not
change the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 1, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: June 15, 1999.
Brief description of amendments: The amendments revised Technical
Specification (TS) 4.7.D.6 by replacing the leakage limit of 11.5
standard cubic feet per hour (scfh) for each main steam isolation valve
(MSIV) with a limit of 46 scfh on the total combined leakage for the
MSIVs of all four main steam lines.
Date of issuance: October 1, 1999.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 175 and 171.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 14, 1999 (64 FR
38024).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 1, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
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: May 19, 1999.
Brief description of amendments: The amendments relocated Technical
Specification 3/4.4.4, ``Chemistry,'' from the TS to the Updated Final
Safety Analysis Report (UFSAR) and to an Administrative Technical
Requirement that has been incorporated into the UFSAR by reference.
Date of issuance: October 1, 1999.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 134 and 119.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 14, 1999 (64 FR
38024).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 1, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Jacobs Memorial Library, 815
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,
Illinois 61348-9692.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: May 11, 1999, as supplemented by
letter dated July 13, 1999.
Brief description of amendments: The amendments revised the
Technical Specifications by incorporating changes to the pressure-
temperature limits; the heatup, cooldown, and inservice test limits for
the reactor coolant system to a maximum of 33 Effective Full Power
Years; the low temperature overpressure protection system; and
operational requirements for the reactor coolant pumps.
Date of Issuance: October 1, 1999.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: Unit 1-307; Unit 2-307; Unit 3-307.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 16, 1999 (64 FR
32289).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 1, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Energy Northwest, Docket No. 50-397, WNP-2, Benton County, Washington
Date of application for amendment: April 7, 1999, as supplemented
by letters dated May 25, June 21, August 2, and August 30, 1999.
Brief description of amendment: The amendment revises the minimum
critical power ratio safety limits.
Date of issuance: September 27, 1999.
Effective date: September 27, 1999.
Amendment No.: 158.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 19, 1999 (64 FR
27329).
The May 25, June 21, August 2 and August 30, 1999, supplemental
letters provided additional clarifying information that did not expand
the scope of the application as originally noticed and did not change
the staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 27, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Energy Northwest, Docket No. 50-397, WNP-2, Benton County, Washington
Date of application for amendment: April 20, 1999, as supplemented
by letter dated September 9, 1999.
Brief description of amendment: The amendment revised Technical
Specification 3.4.11, ``RCS Pressure and Temperature (PT) Limits,'' for
32 effective full power years (EFPY) using the latest vessel beltline
material and fluence data.
Date of issuance: October 6, 1999.
Effective date: October 6, 1999.
[[Page 56540]]
Amendment No.: 159.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 19, 1999 (64 FR
27330).
The September 9, 1999, supplemental letter provided additional
clarifying information, did not significantly expand the scope of the
application as originally noticed and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 6, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: May 14, 1999, as supplemented by letters
dated June 17, and September 7, 15, 17, and 24, 1999.
Brief description of amendment: The amendment revises the Technical
Specification requirements affecting the surveillance criteria for that
portion of the once-through steam generator tubes regarded as a
primary-to-secondary pressure boundary located within the upper
tubesheet and impacted by a specific degradation mechanism, namely,
outside diameter intergranular attack.
Date of issuance: October 4, 1999.
Effective date: As of the date of issuance and shall be implemented
prior to startup from the Unit 1 Cycle 15 refueling outage.
Amendment No.: 202.
Facility Operating License No. DPR-51: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 2, 1999 (64 FR
29709).
The June 17, and September 7, 15, 17, and 24, 1999, letters
provided clarifying and additional information that did not change the
scope of the May 14, 1999, application and the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 4, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: July 2, 1998, as supplemented by letters
dated July 7 and August 24, 1999.
Brief description of amendment: The amendment changes the ACTION
requirements for Technical Specification (TS) 3/4.3.2 for the Emergency
Feedwater Actuation Signal (EFAS). This change revises the allowed
outage time for a channel of EFAS to be in the tripped condition from
``prior to entry into the applicable MODE(S) following the next COLD
SHUTDOWN'' to the more restrictive time limit of 48 hours and adds a
shutdown requirement. Additionally, the TS 3.0.4 exemption is removed
from the ACTION statement for the tripped condition. Changes to TS
Bases Section 3/4.3.2 are also included to support the changes.
Date of issuance: October 6, 1999.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 154.
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 16, 1998 (63
FR 69339). The July 7 and August 24, 1999, letters provided additional
information that did not change the scope of the July 2, 1998,
application and the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 6, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of application for amendment: February 23, 1999.
Brief description of amendment: This amendment removes redundant
boron concentration monitoring requirements specified for Modes 3
through 6 contained in TS 3/4.1.2.9, ``Reactivity Control Systems-Boron
Dilution.''
Date of Issuance: October 4, 1999.
Effective Date: October 4, 1999.
Amendment No.: 104.
Facility Operating License No. NPF-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 25, 1999 (64 FR
46440).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 4, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit 3, Citrus County, Florida
Date of application for amendment: May 5, 1999, as supplemented May
21, May 28, August 20, and September 2, 1999.
Brief description of amendment: Changes the Crystal River Unit 3
Technical Specifications to allow an alternate repair criteria (ARC)
for axial tube end crack-like indications in the upper and lower
tubesheets of the Once-Through Steam Generators (OTSGs). The ARC will
allow leaving OTSG tubes with axially oriented tube end cracks located
within the clad region of the tube-to-tubesheet roll joint in service.
Date of issuance: October 1, 1999.
Effective date: October 1, 1999.
Amendment No.: 188.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 2, 1999 (64 FR
29710). The May 21, May 28, August 20, and September 2, 1999,
supplements did not affect the original no significant hazards
consideration determination, or expand the scope of the amendment
request as originally noticed.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 1, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: December 29, 1998, as
supplemented June 18, 1999.
Brief description of amendment: Transfer of the license for Crystal
River Unit 3, to the extent it is held by the City of Tallahassee, to
Florida Power Corporation.
Date of issuance: October 1, 1999.
Effective date: October 1, 1999.
Amendment No.: 189.
Facility Operating License No. DPR-31: Amendment revised the
License.
[[Page 56541]]
Date of initial notice in Federal Register: February 26, 1999 (64
FR 9544). The supplemental letter dated June 18, 1999, did not change
the original proposed no significant hazards consideration
determination, or expand the scope of the amendment request as
originally noticed.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal River, Florida 34428.
North Atlantic Energy Service Corporation, et al., Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: December 16, 1998.
Brief description of amendment: The amendment relocates Technical
Specification (TS) 3/4.7.10 ``Area Temperature Monitoring,'' and the
associated TS Table 3.7-3, to the Technical Requirements Manual, which
is referenced in the Seabrook Station Updated Final Safety Analysis
Report and is the implementing manual for the TS improvement program
referenced in Section 6.7 of the TSs.
Date of issuance: October 1, 1999.
Effective date: As of the date of issuance, and shall be
implemented within 90 days.
Amendment No.: 63.
Facility Operating License No. NPF-86: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6700).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 1, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833.
North Atlantic Energy Service Corporation, et al., Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: March 27, 1998, as supplemented by
letter dated June 17, 1998.
Brief description of amendment: To revise Technical Specification
(TS) 3.7.6.1, Control Room Emergency Makeup Air and Filtration, and TS
3.7.6.2, Control Room Air Conditioning, to delete the restriction to
suspend all operations involving positive reactivity changes during the
plant conditions specified.
Date of issuance: October 5, 1999.
Effective date: As of its date of issuance, and shall be
implemented within 60 days.
Amendment No.: 64.
Facility Operating License No. NPF-86: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 22, 1998 (63 FR
19973). The June 17, 1998, supplement provided clarifying information
and did not change the staff's proposed no significant hazards
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 5, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: March 31, 1999.
Brief description of amendment: The amendment revised Sections
2.10.4, 3.1, and Table 3-3 of the technical specifications to increase
the minimum required reactor coolant system (RCS) flow rate and change
surveillance requirements for RCS flow rate.
Date of issuance: October 6, 1999.
Effective date: October 6, 1999, to be implemented within 30 days
from the date of issuance.
Amendment No.: 193.
Facility Operating License No. DPR-40. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 19, 1999 (64 FR
27322).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 6, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
PECO Energy Company, Public Service Electric and Gas Company Delmarva
Power and Light Company, and Atlantic City Electric Company, Docket
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2
and 3, York County, Pennsylvania
Date of application for amendments: March 29, 1999, as supplemented
July 21, 1999.
Brief description of amendments: The amendments delete the
surveillance requirement (SR) associated only with the refuel platform
fuel grapple fully retracted position interlock input, which is
currently required by the Peach Bottom Atomic Power Station, Units 2
and 3, Technical Specification SR 3.9.1.1.
Date of issuance: September 24, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendments Nos.: 229 and 232.
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 11, 1999 (64 FR
43774). The July 21, 1999, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 24, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (Regional Depository) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Docket No.
50-278, Peach Bottom Atomic Power Station, Unit No. 3, York County,
Pennsylvania
Date of application for amendment: July 12, 1999, and supplemented
August 30, 1999.
Brief description of amendment: The amendment changed the minimum
critical power ratio safety limit and the approved methodologies
referenced in the core operating limits report.
Date of issuance: October 5, 1999.
Effective date: As of date of issuance and shall be implemented
prior to the start of Peach Bottom Atomic Power Station Unit No. 3,
Cycle 13 operation.
Amendment No.: 233.
Facility Operating License No. DPR-56: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 11, 1999 (64 FR
43777). The August 30, 1999, letter provided additional information but
did not change the initial proposed no significant hazards
consideration determination or expand the amendment beyond the scope of
the initial notice.
[[Page 56542]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 5, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (Regional Depository) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: September 10, 1998 (PCN-496),
as supplemented July 19, 1999.
Brief description of amendments: The amendments delete Technical
Specification 3.6.7 relating to hydrogen recombiners.
Date of issuance: October 7, 1999.
Effective date: October 7, 1999, to be implemented within 30 days
of issuance.
Amendment Nos.: Unit 2--159; Unit 3--150.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 11, 1999 (64 FR
43778).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 7, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P.O. Box 19557, Irvine, California 92713.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: November 6, 1998.
Brief Description of amendments: The amendments revise the TS
nuclear instrumentation system (NIS) surveillance requirements. The
revised TS changes require Southern Nuclear Company to adjust the NIS
power range channels only when calorimetric-calculated power is greater
than the power range indicated power by more than +2 percent rated
thermal power. The proposed TS changes are for both the current TS and
the improved TS.
Date of issuance: October 1, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 144 and 135
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: January 27, 1999 (64 FR
4160).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 1, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of application for amendments: April 13, 1999, as supplemented
by letter dated August 26, 1999.
Brief description of amendments: The amendments revise Technical
Specifications (TS) to update Limiting Condition for Operation (LCO)
3.0.4 and Surveillance Requirements (SR) 3.0.4 in the existing TS to be
consistent with the versions of the LCO 3.0.4 and SR 3.0.4 as they
appear in Revision 1 to NUREG-1431. The proposed change also adds the
words ``or that are part of a shutdown of the unit,'' to LCO 3.0.4 to
allow reactor shutdowns that are not necessarily required by other TS
Required Actions.
Date of issuance: September 30, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1--108; Unit 2--86.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 11, 1999 (64 FR
43779). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 30, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, Georgia.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: July 29, 1999.
Brief description of amendments: The amendments revise TS Section
3.1.7, ``Standby Liquid Control (SLC) System.'' The revision replaces
``greater than the Region B limits,'' which could be misleading, with
``within the Region B limits.''
Date of issuance: September 24, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1--217; Unit 2--158.
Facility Operating License Nos. DPR-57 and NPF-5: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 25, 1999 (64 FR
46449). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 24, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia.
Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear
Plant, Unit 3, Limestone County, Alabama
Date of application for amendment: July 28, 1999 (TS-398).
Brief description of amendment: The amendment revises the Technical
Specifications (TS) to implement operability and surveillance
requirements for the previously-installed Oscillation Power Range
Monitor trip function.
Date of issuance: September 27, 1999.
Effective date: As of the date of issuance, to be implemented at
the end of the Cycle 9 outage.
Amendment No.: 221.
Facility Operating License No. DPR-68: Amendment revises the TS.
Date of initial notice in Federal Register: August 25, 1999 (64 FR
46450). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 27, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: February 26, 1999 (TS 98-08).
Brief description of amendments: The amendments relocate Sequoyah
Nuclear
[[Page 56543]]
Plant Technical Specification (TS) 3.7.6, ``Flood Protection Plan,''
and its associated bases from the TS to the Technical Requirements
Manual. Future changes to the Flood Protection Plan will be processed
in accordance with 10 CFR 50.59.
Date of issuance: October 6, 1999.
Effective date: As of the date of issuance to be implemented no
later than 45 days after issuance.
Amendment Nos.: 247 and 238.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the TS.
Date of initial notice in Federal Register: March 24, 1999 (64 FR
14286) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 6, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: July 20, 1999, as supplemented
August 13, 1999.
Brief description of amendment: The amendment modifies the
operability requirements for the high pressure cooling systems--High
Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling
(RCIC), and Automatic Depressurization System (ADS)--and the safety and
relief valves, and adds a time limitation for conducting operability
testing of HPCI and RCIC.
Date of Issuance: October 1, 1999.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 177
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 31, 1999 (64 FR
47537)
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated October 1, 1999.
No significant hazards consideration comments received: No
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear
Power Station, Vernon, Vermont
Date of application for amendment: June 29, 1999 Brief description
of amendment: The amendment revises the leak rate requirements for the
main steam line isolation valves. Specifically, a total allowable
leakage rate for the sum of the four main steam lines is established
that is equal to four times the current allowable individual main steam
line isolation valve leakage rate. The allowable individual main steam
line isolation valve leakage rate is revised to be one half of the
allowable total leakage rate.
Date of Issuance: October 1, 1999.
Effective date: 10/01/99, and shall be implemented within 30 days.
Amendment No.: 178
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 28, 1999 (64 FR
40909).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated October 1, 1999.
No significant hazards consideration comments received: No
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
For the Nuclear Regulatory Commission.
Dated at Rockville, Maryland, this 13th day of October, 1999.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 99-27210 Filed 10-19-99; 8:45 am]
BILLING CODE 7590-01-P
This is a copy of a public record, reproduced as it was published. It is not legal advice, and it may not be the version a court would rely on. Check the official source before you cite it.