[Federal Register Volume 64, Number 66 (Wednesday, April 7, 1999)]
[Notices]
[Pages 17021-17040]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-8503]
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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 Pub. L. 97-415, the U.S. Nuclear Regulatory Commission
(the Commission or NRC staff) is publishing this regular biweekly
notice. Pub. L. 97-
[[Page 17022]]
415 revised section 189 of the Atomic Energy Act of 1954, as amended
(the Act), to require the Commission to publish notice of any
amendments issued, or proposed to be issued, under a new provision of
section 189 of the Act. This provision grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from March 13, 1999, through March 26, 1999. The
last biweekly notice was published on March 24, 1999 (64 FR 14278).
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 April 23, 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
[[Page 17023]]
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: 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 Nos.
1, 2, and 3, Maricopa County, Arizona
Date of amendments request: February 26, 1999.
Description of amendments request: The proposed amendment would
revise Technical Specification (TS) 3.5.3, ``Emergency Core Cooling
System--Operating,'' to extend the completion time for one inoperable
low pressure safety injection subsystem from 72 hours to 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. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed amendment will extend the Completion Time
for one inoperable low pressure safety injection (LPSI) subsystem in
Technical Specification (TS) 3.5.3, Emergency Core Cooling Systems
(ECCE)[S]--Operating, from 72 hours to 7 days. The LPSI subsystem is
part of the ECCS train and part of the shutdown cooling subsystem.
The LPSI components are not accident initiators in any accident
previously evaluated. Therefore, this change does not involve a
significant increase in the probability of an accident previously
evaluated.
The LPSI system is primarily designed to mitigate the
consequences of a large break loss of coolant accident (LOCA). These
proposed changes do not affect any of the assumptions used in the
deterministic LOCA analysis.
In order to evaluate the LPSI Completion Time extension with
respect to the ECCS, probabilistic safety analysis (PSA) methods
were utilized. The results of these analyses show no significant
increase in the core damage frequency. As a result, there would be
no significant increase in the consequences of an accident
previously evaluated. These analyses are detailed in CE NPSD-995,
Combustion Engineering Owners Group ``Joint Applications Report for
Low Pressure Safety Injection System AOT Extension,'' May 1995, as
supplemented by updated PVNGS data provided in the attachment to
this enclosure.
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 does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed amendment will extend the Completion Time for one
inoperable low pressure safety injection (LPSI) subsystem in
Technical Specification (TS) 3.5.3, Emergency Core Cooling Systems
(ECCE)[S]--Operating, from 72 hours to 7 days. The proposed change
does not change the design, configuration, or method of operation of
the plant. Therefore, this change does not create the possibility of
a new or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change does not involve a significant reduction in
a margin of safety. The proposed amendment will extend the
Completion Time for one inoperable low pressure safety injection
(LPSI) subsystem in Technical Specification (TS) 3.5.3, Emergency
Core Cooling Systems (ECCE)[S]--Operating, from 72 hours to 7 days.
The proposed change does not affect the limiting conditions for
operation or their bases used in the deterministic analyses to
establish the margin of safety. PSA evaluations were used to
evaluate these changes. These evaluations demonstrate that the
changes will be risk neutral or risk beneficial for PVNGS. These
evaluations are detailed in CE NPSD-995, as supplemented by updated
data provided in the attachment to this enclosure.
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
amendments request involve 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 Project Director: William H. Bateman.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: January 22, 1999.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) Sections 3.7.D.1.g, 6.2.2.h and
6.3.1. Specifically, (1) Section 3.7.D.1.g would be revised to correct
an editorial error; (2) Section 6.2.2.h would be revised to change the
senior reactor operator license requirement for the Operations Manager;
and (3) Section 6.3.1 would modify the qualification requirement for
the Operations Manager.
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 [to Section 3.7.D.1.g] is administrative in
nature. It involves making an editorial change to provide the
correct functional description of the breakers. This change does not
affect possible initiating events for accidents previously evaluated
or alter the configurations or operation of the facility. The
Limiting Safety Systems Settings and Safety Limits specified in the
current Technical Specifications
[[Page 17024]]
remain unchanged. Therefore, the proposed change to the subject
Technical Specification would not increase the probability or
consequences of an accident previously evaluated.
The proposed change [to Section 6.2.2.h] is administrative in
nature. The individual who provides the day to day direction of the
activities of the operating shift will still possess an SRO [Senior
Reactor Operator] license and this proposed change is consistent
with the statement in NUREG-1431, Section 5.2.2.f. This change does
not affect possible initiating events for accidents previously
evaluated or alter the configuration or operation of the facility.
The Limiting Safety Systems Settings and Safety Limits specified in
the current Technical Specifications remain unchanged. Therefore,
the proposed change to the subject Technical Specification would not
increase the probability or consequences of an accident previously
evaluated.
The proposed change [to Section 6.3.1] is administrative in
nature. The individual who provides the day to day direction of the
activities of the operating shift will still possess an SRO license
and this proposed change is consistent with the statement in NUREG-
1431, Section 5.2.2.f. This change does not affect possible
initiating events for accidents previously evaluated or alter the
configuration or operation of the facility. The Limiting Safety
Systems Settings and Safety Limits specified in the current
Technical Specifications remain unchanged. Therefore, the proposed
change to the subject Technical Specification would not increase 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.
As stated above, the proposed change [to Section 3.7.D.1.g] is
administrative in nature. The safety analysis of the facility
remains complete and accurate. There are no physical changes to the
facility and the plant conditions for which the design basis
accidents have been evaluated are still valid. The operating
procedures and emergency procedures are unaffected. Consequently, no
new failure modes are introduced as a result of the proposed change.
Therefore, the proposed change will not initiate any new or
different kind of accident.
The proposed change [to Section 6.2.2.h] is administrative in
nature. The safety analysis of the facility remains complete and
accurate. There are no physical changes to the facility and the
plant conditions for which the design basis accidents have been
evaluated are still valid. The operating procedures and emergency
procedures are unaffected. Consequently, no new failure modes are
introduced as a result of the proposed changes. Therefore, the
proposed change will not initiate any new or different kind of
accident.
The proposed change [to Section 6.3.1] is administrative in
nature. The safety analysis of the facility remains complete and
accurate. There are no physical changes to the facility and the
plant conditions for which the design basis accidents have been
evaluated are still valid. The operating procedures and emergency
procedures are unaffected. Consequently, no new failure modes are
introduced as a result of the proposed changes. Therefore, the
proposed change will not initiate any new or different kind of
accident.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change [to Section 3.7.D.1.g] is administrative in
nature. Since there are no changes to the operation of the facility
or physical design the Updated Final Safety Analysis Report (UFSAR)
design basis, accident assumptions, or Technical Specification Bases
are not affected. Therefore, the proposed changes will not result in
a reduction in the margin of safety.
The proposed change [to Section 6.2.2.h] is administrative in
nature. Since there are no changes to the operation of the facility
or physical design the Updated Final Safety Analysis Report (UFSAR)
design basis, accident assumptions, or Technical Specification Bases
are not affected. Therefore, the proposed changes will not result in
a reduction in the margin of safety.
The proposed change [to Section 6.3.1] is administrative in
nature. Since there are no changes to the operation of the facility
or physical design the Updated Final Safety Analysis Report (UFSAR)
design basis, accident assumptions, or Technical Specification Bases
are not affected. Therefore, the proposed changes will not result in
a 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 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: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: S. Singh Bajwa, Director.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: January 22, 1999.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) Section 4.3. Specifically, the
revision would permit the reactor coolant system (RCS) leak test to be
performed at normal operating pressure after it has been closed
following normal opening in lieu of a hydrostatic test being performed
at 2335 psig.
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 license amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The change proposes a system leakage test for
the RCS that is comparable to the hydrostatic test that it replaces,
as acknowledged by the NRC approval of ASME Code Case N-498,
``Alternative Rules for 10-Year Hydrostatic Pressure Testing for
Class 1 and 2 Systems Section XI, Division 1,'' and the ASME
[American Society for Mechanical Engineers] Boiler and Pressure
Vessel Code, Section XI. [. . .] The proposed change to substitute a
system leak test at normal operating pressure in lieu of the
hydrostatic test at 2335 psig will minimize challenge to plant
safety and demonstrate leak tightness of the RCS. Therefore, the
proposed change would not involve a 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 license amendment does not create the possibility
of a new or different kind of accident from any accident previously
evaluated. The proposed changes do not involve the addition of any
new or different type of equipment, nor do they involve the
operation of equipment required for safe operation of the facility
in a manner different from those addressed in the Updated Final
Safety Analysis Report. [. . .] Based on industry experience, it is
expected that any leaks would be discovered by the leak test at
normal operating pressure.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed license amendment does not involve a significant
reduction in a margin of safety. The proposed changes do not
adversely affect performance of any safety related system or
component, instrument operation, or safety system setpoints and do
not result in increased severity of any of the accidents considered
in the safety analysis. Although the current basis states that if
the system does not leak at 2335 psig (operating pressure + 100
psig) it will be leak tight during normal operation, industry
experience demonstrates that leaks are not discovered as a result of
hydrostatic test pressure propagating a preexisting flaw through
wall. In most cases, leaks are discovered when the system is at
normal operating pressure. Also, testing will continue to be
performed as required by the ASME Boiler and Pressure Vessel Code
Section XI.
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.
[[Page 17025]]
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: S. Singh Bajwa, Director.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: March 8, 1999.
Description of amendment request: The proposed amendments would
delete certain requirements from Technical Specification (TS) Section
6.0 ``Administrative Controls'' that are adequately controlled by
existing regulations, other than 10 CFR 50.36 and the TS. The
amendments also relocate selected requirements from TS Section 6.0 to
the licensee's controlled documents such as the Turkey Point Units 3
and 4 Updated Final Safety Analysis Report (UFSAR). The amendments also
clarify certain provisions of TS Section 6.0. The proposed changes are
to relocate, revise, delete, or clarify the following provisions of the
TS:
------------------------------------------------------------------------
Existing TS section Subject Proposed change
------------------------------------------------------------------------
6.2.2.f................ Administrative Controls Partly delete, partly
on Working Hours of relocate within TS.
Plant Staff.
Table 6.2-1............ Minimum Shift Crew Clarify.
Composition.
6.2.3.................. Shift Technical Advisor Clarify.
6.4.................... Training............... Delete.
6.5.................... Review and Audit....... Relocate to UFSAR.
6.6.................... Reportable Event Action Partly delete, partly
relocate to UFSAR.
6.8.2.................. Review and Approval of Relocate to UFSAR.
Procedures.
6.8.3.................. Temporary Changes to Relocate to UFSAR.
Procedures.
6.8.4.b................ In-Plant Radiation Relocate to UFSAR.
Monitoring.
6.8.4.g................ Radiological Relocate to UFSAR.
Environmental
Monitoring Program.
6.10................... Record Retention....... Relocate to UFSAR.
6.11................... Radiation Protection Relocate to UFSAR.
Program.
6.12................... High Radiation Area.... Clarify.
6.13................... Process Control Program Relocate to UFSAR.
(PCP).
6.14................... Offsite Dose Revise to reflect
Calculation Manual changes to 6.5 &
(ODCM). 6.10.
------------------------------------------------------------------------
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of the plant in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because the proposed changes are administrative in nature. These
proposed changes will not involve a significant increase in the
probability or consequences of an accident previously evaluated
because they do not affect assumptions contained in plant safety
analyses, the physical design and/or operation of the plant, nor do
they affect Technical Specifications that preserve safety analysis
assumptions. None of the proposed changes involve a physical
modification to the plant, a new mode of operation or a change to
the UFSAR transient analyses. No Limiting Condition for Operation,
ACTION statement or Surveillance Requirement is affected by any of
the proposed changes. Also, these proposed changes, in themselves,
do not reduce the level of qualification or training such that
personnel requirements would be decreased. Further, the Proposed
changes do not alter the design, function, or operation of any plant
component. Therefore, the proposed changes do not affect the
probability or consequences of accidents previously evaluated.
2. Operation of the plant in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The changes being proposed are administrative in nature and do not
affect assumptions contained in plant safety analyses, the physical
design and/or modes of plant operation defined in the plant
operating license, or Technical Specifications that preserve safety
analysis assumptions. The proposed changes do not introduce a new
mode of plant operation or surveillance requirement, nor involve a
physical modification to the plant. The proposed changes are
administrative in nature. The changes propose to revise, delete, or
relocate the stated administrative control provisions from the TS to
the UFSAR whereby adequate control of information is maintained.
Furthermore, the proposed changes do not alter the design, function,
or operation of any plant components. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Operation of the plant in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed changes do not involve a significant reduction in a
margin of safety because they are administrative in nature. The
operating limits and functional capabilities of the affected
systems, structures, and components are unchanged by the proposed
amendments. None of the proposed changes involve a physical
modification to the plant, a new mode of operation or a change to
the UFSAR transient analyses. No Limiting Condition for Operation,
ACTION statement, or Surveillance Requirement is affected.
Additionally, the proposed changes do not alter the scope of
equipment currently required to be OPERABLE or subject to
surveillance testing, nor does the proposed change affect any
instrument setpoints or equipment safety functions. Therefore, the
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 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: Florida International
University, University Park, Miami, Florida 33199.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Project Director: Cecil O. Thomas.
[[Page 17026]]
GPU Nuclear, Inc. etal., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of amendment request: December 23, 1998.
Description of amendment request: The proposed Technical
Specification (TS) change request will change the surveillance
frequency for verifying the operability of motor-operated isolation
valves and condensate makeup valves in the Isolation Condenser TS
4.8.A.1 and Bases page from once per month to once per 3 months.
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 a[n] 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 Isolation Condenser motor-operated
isolation valves and condensate makeup valves from once per month to
once per three months. The proposed change does not alter the actual
surveillance requirements, the limits and restrictions 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 Project Director: Elinor G. Adensam.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of amendment request: February 12, 1999.
Description of amendment request: The proposed Technical
Specification (TS) change will delete the organizational chart and the
related organizational references from the Appendix B Environmental TS
and revise the appearance and format of the Environmental TS.
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 because deletion of the organization charts and other
organizational references in the [Environmental Technical
Specifications] ETS does not affect plant operation. GPU Nuclear
will continue to inform the NRC of organizational changes through
other required controls.
2. The proposed change does not create the possibility of a new
or different type of accident than previously evaluated because the
proposed change is administrative in nature, and no physical
alteration of plant configuration, changes to setpoints or operating
parameters are proposed.
3. The proposed change does not involve a significant reduction
in the margin of safety because it does not alter the design,
function or manner of operation of any structures, systems or
components. Organizational structure or its representation does not
directly impact 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.
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 Project Director: Elinor G. Adensam.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn
County, Iowa.
Date of amendment request: February 18, 1999.
Description of amendment request: The proposed amendment would
revise Duane Arnold Energy Center (DAEC) Technical Specification (TS)
Table 3.3.6.1-1, ``Primary Containment Isolation Instrumentation,'' by
deleting the manual initiation function of the high pressure coolant
injection (HPCI) system and reactor core isolation cooling (RCIC)
system isolation. A related condition as well as corresponding
surveillance requirements and bases would also be deleted. Thus, the
change would (1) revise Table 3.3.6.1-1 by removing items 3j. and 4.j.;
(2) revise Note 2 to Surveillances to Licensing Condition for Operation
(LCO) 3.3.6.1 by deleting information regarding items 3 j. and 4.j.;
and (3) revise LCO 3.3.6.1 by removing Condition G and Surveillance
Requirement 3.3.6.1.10.
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:
After reviewing this proposed amendment, we [the licensee] have
concluded:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The Manual Initiation Function for HPCI and
RCIC Isolation is not considered to be an initiator for any accident
previously evaluated in the UFSAR. Therefore, this change does not
involve a significant increase in the probability of any previously
evaluated accidents. The Manual Initiation push button channels
introduce signals into HPCI and RCIC System isolation logics that
are redundant to the automatic protective instrumentation and
provide manual isolation capability only if a system initiation
signal is present. Technical Specification Section 3.3.6.1 Condition
G requires isolation of the System flowpath, which renders the
System inoperable and reduces the availability of the System due to
the failure of a manually initiated isolation, an isolation which is
not assumed in any transient or accident analysis in the UFSAR.
Removal of the Manual Initiation Function
[[Page 17027]]
for HPCI and RCIC from the Primary Containment Isolation
Instrumentation Section of Technical Specifications does not affect
the automatic protective instrumentation and the automatic isolation
capability. Therefore, this change does not significantly increase
the consequences of a previously analyzed accident.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. The proposed change introduces no new mode of plant
operation and does not involve physical modification to the plant.
Therefore, it does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. The proposed amendment will not involve a significant
reduction in a margin of safety. The proposed change deletes the
Manual Initiation Function from Technical Specifications, but no
significant reduction in a margin of safety is involved. Technical
Specification Section 3.3.6.1 Condition G requires isolation of the
System flowpath, which renders the System inoperable and reduces the
availability of the System due to the failure of a manually
initiated isolation, an isolation that is not assumed in any
transient or accident analysis in the UFSAR. Removal of the Manual
Initiation Function for HPCI and RCIC from the Primary Containment
Isolation Instrumentation Section of Technical Specifications does
not affect the automatic protective instrumentation and the
isolation capability. This change is acceptable based on the fact
that the Manual Initiation Function is not assumed in any accident
or transient analysis in the UFSAR.
Based upon the above, we [licensee] have determined that the
proposed amendment will not involve a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, IA 52401.
Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Project Director: T.J. Kim, Acting.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: March 1, 1999.
Description of amendment request: The proposed amendment would
change the Cooper Nuclear Station (CNS) Technical Specifications (TSs)
to revise the calibration frequency of the reactor recirculation flow
transmitters from once every 184 days to once every 18 months. This
calibration is required as part of TS Surveillance Requirement (SR)
3.3.1.1.10.
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 will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Changing the calibration frequency of the recirculation
loop flow transmitters from 184 days to 18 months may increase the
amount of drift experienced by the transmitters. However, CNS
calculation (NEDC 98-024 [forwarded by letter dated March 10, 1999])
takes into account the 18 month calibration intervals. This
calculation, performed in accordance with the General Electric (GE)
setpoint methodology for CNS, demonstrates that the expected drift
is not significant, and is consistent with past operating
experience. Changing the calibration frequency of the flow
transmitters does not change any of the precursors assumed in the
accident analysis. Therefore, changing the calibration frequency for
flow transmitters from 184 days to 18 months does not involve a
significant increase in the probability of an accident previously
evaluated in the USAR [Updated Safety Analysis Report].
The proposed change will not create the possibility of a new or
different kind of accident than evaluated in the USAR. The proposed
change does not result in any physical change to plant structures,
systems, or components. The proposed change does not alter the form,
fit, or function of any equipment or components credited in the
accident analyses described in the USAR. Therefore, changing the
test frequency does not create the possibility of a new or different
kind of accident.
The proposed change will not involve a significant reduction in
a margin of safety. This conclusion is based on the fact that the
proposed change is consistent with the drift assumptions used in CNS
approved calculation (NEDC 98-024). The calibration frequency of 18
months is consistent with the operating practices prior to
conversion to Improved Technical Specifications, and is consistent
with past operating practice at CNS. Therefore, changing the
calibration frequency from 184 days to 18 months 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: Auburn Memorial Library, 1810
Courthouse Avenue, Auburn, Nebraska 68305.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, Nebraska 68602-0499.
NRC Project Director: George Dick, Acting.
Northeast Nuclear Energy Company (NNECO), et al., Docket Nos. 50-336
and 50-423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New
London County, Connecticut
Date of amendment request: March 5, 1999.
Description of amendment request: The proposed amendment would
relocate certain Technical Specification (TS) Section 6.0
administrative controls to the NRC-approved Northeast Utilities Quality
Assurance Program (NUQAP) Topical Report. Specifically, Sections 6.2.3
(Unit 3 only), 6.5, 6.6 (partial), 6.7 (partial), and 6.10. The
proposed amendment would also delete parts of Section 6.6 and 6.7
because their requirements are duplicated in existing regulations or
elsewhere in the TS. In addition, the proposed amendment would modify
the table of contents and other TS sections to incorporate the
aforementioned changes (e.g., correct references).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
In accordance with 10 CFR 50.92, NNECO has reviewed the attached
proposed changes and has concluded that they do not involve a
Significant Hazards Consideration (SHC). The basis for this
conclusion is that the three criteria of 10 CFR 50.92 are not
compromised. The proposed changes are not a SHC because the proposed
change will not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
No design basis accidents are affected by these proposed
changes. The proposed changes relocate portions of the Technical
Specifications to the NUQAP Topical Report or remove duplicate
sections and are being proposed to eliminate the need for a T.S.
change each time there is a related change in the administrative
controls for the site.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
There are no changes in the way the plant is operated due to
these revisions. The potential for an unanalyzed accident is not
created. There is no impact on plant response, and no new failure
modes are introduced. The proposed deletions and
[[Page 17028]]
editorial changes have no impact on safety limits or design basis
accidents, and have no potential to create a new or unanalyzed
event.
3. Involve a significant reduction in a margin of safety.
These changes do not directly affect any protective boundaries
nor do they impact the safety limits for the protective boundaries.
These proposed changes relocate portions of the administrative
controls to the NUQAP Topical Report or are editorial in nature.
Therefore, there is no reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Project Director: Elinor G. Adensam.
PECO Energy Company, Docket No. 50-353, Limerick Generating Station,
Unit 2, Montgomery County, Pennsylvania
Date of amendment request: March 11, 1999.
Description of amendment request: The proposed revision to the
Technical Specifications (TSs) involves a change to TS Section 2.1 and
its associated TS Bases to revise the minimum critical power ratio
(MCPR) Safety Limits for Cycle 6.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The revised MCPR Safety Limits for LGS Unit 2 Technical
Specifications, and their use to determine cycle-specific thermal
limits, have been calculated using NRC-approved methods (i. e,
GESTAR-II, Rev. 13) and are based on LGS, Unit 2, Cycle 6 specific
inputs. The use of these methods assures that the SLMCPR [safety
limit minimum critical power ratio] value is within the existing
design and licensing basis, and cannot increase the probability or
severity of an accident.
The basis for the MCPR Safety Limit calculation is to ensure
that greater than 99.9 percent of all fuel rods in the core avoid
transition boiling if the limit is not violated. The MCPR Safety
Limit preserves the existing margin to transition boiling and fuel
damage in the event of a postulated accident. The probability of
fuel damage is not increased.
Therefore, the proposed TS changes do not involve an increase in
the probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The MCPR Safety Limit is a Technical Specification numerical
value designed to ensure that fuel damage from transition boiling
does not occur as a result of the limiting postulated accident. The
MCPR Safety Limit is not an accident initiator; therefore, it cannot
create the possibility of any new type of accident. The new MCPR
Safety Limits are calculated using NRC-approved methods (i.e.,
GESTAR-II, Rev. 13) and are based on LGS, Unit 2, Cycle 6 specific
inputs.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed TS changes do not involve a significant
reduction in the margin of safety.
The margin of safety as defined in the TS Bases will remain the
same. The new MCPR Safety Limits are calculated using NRC-approved
methods (i.e., GESTAR-II, Rev. 13), which are in accordance with the
current fuel design and licensing criteria, and are based on LGS,
Unit 2, Cycle 6 specific inputs. The MCPR Safety Limit remains high
enough to ensure that greater than 99.9 percent of all fuel rods in
the core will avoid transition boiling if the limit is not violated,
thereby preserving the fuel cladding integrity.
Therefore, the proposed TS changes do not involve a 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.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Project Director: Elinor G. Adensam.
PP&L, Inc., Docket No. 50-387, Susquehanna Steam Electric Station, Unit
1, Luzerne County, Pennsylvania
Date of amendment request: March 12, 1999.
Description of amendment request: The amendment would modify the
Susquehanna Steam Electric Station, Unit 1, Technical Specifications
Table 3.3.5.1-1 ``Emergency Core Cooling System Instrumentation.'' The
change updates the allowable values for both the Core Spray (CS) and
Low Pressure Coolant Injection System (LPCI) ``Reactor Steam Dome
Pressure--Low'' functions for initiation and injection permissive.
Specifically, the allowable values are being changed from a specified
minimum pressure to a specified allowable pressure band. This more
restrictive allowable value range will prevent CS and LPCI system
overpressurization while still permitting injection to prevent fuel
clad temperature limits from being exceeded.
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.
This proposal does not involve an increase in the probability or
consequences of an accident previously evaluated. The proposed
amendment changes the ``Reactor Steam Dome Pressure-Low'' Allowable
Values so to provide further assurance that the Core Spray and RHR
systems will perform their LOCA [Loss-of-coolant accident] design
basis function.
The functional design basis of the Core Spray and LPCI is to
inject water into the reactor vessel to cool the core during a LOCA
by opening the Core Spray and LPCI injection valves when reactor
pressure drops below the reactor vessel low pressure permissive. The
upper analytical limit for the permissive is the Core Spray and LPCI
systems' maximum design pressure, and the lower analytical limit is
the lowest pressure which allows injection to prevent exceeding the
fuel cladding temperature limit. The new allowable values were
selected to lie within the upper and lower limits to ensure there
will be no change in the required logic or functions of the Core
Spray and LPCI systems. These new values do not affect the LOCA or
its ``limiting fault'' frequency of occurrence and do not introduce
any new accidents or malfunctions of equipment important to safety.
Since they do not affect the LOCA, they do not change the
probability of occurrence of the LOCA. The new allowable values do
not change the logic or function of the reactor vessel low pressure
permissive. These new values simply provide the basis for which the
associated pressure
[[Page 17029]]
instruments are to be set to ensure proper operation of Core Spray
and LPCI within the design pressures as described above. Therefore,
the change in allowable values does not increase the probability of
occurrence or the consequences of an accident or malfunction of
equipment important to safety.
Based upon the analysis presented above, PP&L [PP&L, Inc.]
concludes that the proposed action does not involve an 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.
This proposal does not create the probability of a new or
different type of accident from any accident previously evaluated.
The new allowable values do not change any plant systems,
structures, or components, nor do they change any existing or create
any new Core Spray and LPCI logic or functions. The new allowable
values were selected to ensure the required operation of the Core
Spray and LPCI systems within the design pressures described above.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The change does not involve a reduction in the margin of safety.
Technical Specification Bases Section B3.3.5.1 9 (ECCS
Instrumentation) identifies that the low reactor steam dome pressure
signals are used as permissives for operation of the low pressure
ECCS subsystems. The new allowable values were selected so to not
impact the logic, redundancy, operability or surveillance
requirements for these subsystems. The new allowable values maintain
the margin requirements that the Core Spray and LPCI system
pressures such that they do not exceed their system maximum design
pressures and that system pressures are high enough to ensure that
the ECCS injection prevents the fuel peak cladding temperature from
exceeding the limits of 10CFR50.46.
The margin of safety is unaffected by the proposed changes.
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 proposed to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PP&L, Inc., 2 North Ninth St., GENTW3, Allentown, PA 18101-
1179.
NRC Project Director: Elinor G. Adensam.
PP&L, Inc., Docket No. 50-387, Susquehanna Steam Electric Station, Unit
1, Luzerne County, Pennsylvania
Date of amendment request: March 12, 1999.
Description of amendment request: This proposed amendment would
revise the minimum critical power ratio safety limit in Technical
Specification (TS) Section 2.1.1.2. Also, the proposed amendment would
modify the references in TS Section 5.6.5 in order to include only
those references that directly support the generation of the Core
Operating Limit and to remove the reference for the Lead Use
Assemblies, which will be discharged during the next Unit 1 refueling
outage.
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 applicable sections of the FSAR are Chapters 4.4 and 15.
FSAR Chapter 4.4 describes the MCPR Safety Limit, and Chapter 15
describes the transient and accident analyses. The reference to be
added to Section 5.6.5 of the Unit 1 Technical Specifications
describes a NRC approved critical power correlation for
ATRIUMTM-10 fuel. This correlation is appropriate for use
in conservative methodologies for generating MCPR Safety Limits and
MCPR Operating Limits to assure safe operation of Unit 1 with
ATRIUMTM-10 fuel. A discussion of the impact of the
proposed Technical Specification change is provided below.
The proposed change in critical power correlation does not
physically affect the plant or its systems. Thus, it does not
increase the probability of an accident previously evaluated.
A Unit 1 Cycle 12 MCPR Safety Limit analysis was performed for
PP&L by SPC. This analysis used NRC approved methods described in
ANF-524(P)(A), Revision 2 and Supplement 1 Revision 2. These methods
will be used each cycle to calculate the Unit 1 Safety Limits. For
Unit 1 Cycle 12, the critical power performance of the 9 x 9-2 and
ATRIUMTM-10 fuel was determined using the NRC approved
ANFB and ANFB-10 correlations, respectively. The SAFETY LIMIT MCPR
calculations statistically combine uncertainties on feedwater flow,
feedwater temperature, core flow, core pressure, core power
distribution, and uncertainties in the Critical Power Correlation.
The SPC analysis used cycle specific power distributions and
calculated MCPR values such that at least 99.9% of the fuel rods are
expected to avoid boiling transition during normal operation or
anticipated operational occurrences. The resulting two-loop and
single-loop MCPR Safety Limits are included in the proposed
Technical Specification change. Thus, the cladding integrity and its
ability to contain fission products are not adversely affected.
Analyses of the Single Loop Pump Seizure accident with the NRC
approved ANFB-10 correlation for ATRIUMTM-10 fuel
(Reference 1) will be performed to demonstrate that the NRC
acceptance criterion (i.e., small fraction of 10CFR100 dose limits)
is met. Analyses will also be performed to validate the conclusion
that two-loop transients are more severe than those events analyzed
in single-loop operation.
Changes to Section 2.1.1.2 reflect the change from a flow
dependent MCPR Safety Limit to a single value MCPR Safety Limit for
two-loop operation and single-loop operation.
Changes to Reference 5.6.5 delete the methodology used for
critical power analyses for ATRIUMTM-10 fuel and add the
NRC approved ANFB-10 methodology to the list of approved
methodologies. Other changes in Reference 5.6.5 are administrative
in nature because they delete references not directly related to the
generation of Core Operating Limits. No new analysis approaches are
used due to these changes.
Changes to BASES Sections 2.1.1 and 3.2.2 reflect the inclusion
of the ANFB-10 critical power correlation. The range of the
applicability of the ANFB-10 is valid for pressures > 571 psia and
bundle mass fluxes > 0.115 x 10\6\ lb/hr-ft \2\. These values
assure that a valid CPR calculation will result at or above 25% of
rated core thermal power, that is, reactor steam dome pressure
785 psig and core flow 10 Mlbm/hr.
Changes to BASES Sections 3.2.1, 3.2.2, 3.2.3, and 3.2.4 reflect
the removal of Reference 7 for the ABB LUAs, since the four LUAs
will be discharged from Unit 1 during the Unit 1 11th Refueling and
Inspection Outage.
The consequences of transients and accidents will remain within
the criteria approved by the NRC. The methodology used to perform
the analyses has been previously approved by the NRC. Thus, analysis
results using the new methodology will continue to provide assurance
that the reactor will perform its design safety function during
normal operation and design basis events. Therefore, the proposed
action does not involve an 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 changes to the Unit 1 Technical Specifications
(MCPR Safety Limits, removal of methodology references not directly
supporting the generation of Core Operating Limits, removal of the
two references describing previously approved methodology for
applying ANFB to ATRIUMTM-10 fuel, removal of the ABB LUA
reference, and inclusion of the ANFB-10 correlation reference) do
not require any physical plant modifications, physically affect any
plant components, or entail changes in plant operation. Removal of
the Unit 1 Cycle 11 footnote allows Unit 1 Cycle 12 and future cycle
operation with NRC
[[Page 17030]]
approved methodology. Thus, the proposed change does not create the
possibility of a previously unevaluated operator error or a new
single failure. The consequences of transients and accidents will
remain within the criteria approved by the NRC. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The applicable Technical Specification Sections include 2.1.1.2
and 5.6.5.
The changes to the Unit 1 Technical Specifications discussed in
Item 1 above do not require any physical plant modifications,
physically affect any plant components, or entail changes in plant
operation. Therefore, the proposed change will not jeopardize or
degrade the function or operation of any plant system or component
governed by Technical Specifications. The consequences of transients
and accidents will remain within the criteria approved by the NRC.
The proposed MCPR Safety Limits and use of the ANFB-10 critical
power correlation described in the reference added to Section 5.6.5
do not involve a significant reduction in the margin of safety as
currently defined in the Bases of the applicable Technical
Specification sections.
Therefore, the proposed change 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.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Attorney for licensee: Bryan A. Snapp, Esquire, PP&L, Inc., 2 North
Ninth St., Allentown, PA 18101.
NRC Project Director: Elinor G. Adensam.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: March 2, 1999.
Description of amendment request: The proposed amendment would
clarify the use of a ``check valve with flow through the valve
secured'' as a means to isolate an affected containment penetration
(i.e., a penetration with an inoperable penetration barrier) in
Technical Specification 3.6.3 Action b.
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 or consequences of an accident previously evaluated?
The proposed change does not involve an increase in the
probability or consequences of an accident previously evaluated. The
proposed change does not involve any hardware changes. The proposed
change will clarify Technical Specification 3.6.3 Action b to allow
the use of a check valve with the flow through the valve secured as
a means to isolate an inoperable containment penetration. This
change is consistent with the changes identified in NUREG-1431,
``Improved Standard Technical Specifications for Westinghouse
Plants'', Specification 3.6.3 (Containment Isolation Valves), which
identifies check valves with flow through the valve secured as a
type of deactivated automatic valve, and with 10 CFR 50 Appendix A
General Design Criteria 55 and 56, which include the use of check
valves as ``automatic isolation valves''. The proposed change will
not affect the containment isolation valve OPERABILITY requirements
or associated isolation time limits established in the
Specifications. Therefore the proposed change will not affect any
safety margin or safety limit applicable to the facility. Therefore
no increase in the probability or consequences of any accident
previously evaluated will occur.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change will clarify Technical Specification 3.6.3
Action b to allow the use of a check valve with the flow through the
valve secured as a means to isolate an inoperable containment
penetration. The proposed change will not involve any physical
change to plant systems, structures, or components (SSC). This
change is consistent with the changes identified in NUREG-1431,
``Improved Standard Technical Specifications for Westinghouse
Plants'', Specification 3.6.3 (Containment Isolation Valves), which
identifies check valves with flow through the valve secured as a
type of deactivated automatic valve, and with 10 CFR 50 Appendix A
General Design Criteria 55 and 56, which include the use of check
valves as ``automatic isolation valves''. The proposed change only
provides clarification to the existing Specification 3.6.3, and will
not affect the established containment isolation valve OPERABILITY
requirements or associated isolation time limits. Since the proposed
change does not impact operation of the facility as presently
approved, no possibility exists for a new or different kind of
accident from those previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change will clarify Technical Specification 3.6.3
Action b to allow the use of a check valve with the flow through the
valve secured as a means to isolate an inoperable containment
penetration. This change is consistent with the changes identified
in NUREG-1431, ``Improved Standard Technical Specifications for
Westinghouse Plants'', Specification 3.6.3 (Containment Isolation
Valves), which identifies check valves with flow through the valve
secured as a type of deactivated automatic valve, and with 10 CFR 50
Appendix A General Design Criteria 55 and 56, which include the use
of check valves as ``automatic isolation valves''. The proposed
change only provides clarification to the existing Specification
3.6.3, and will not affect the established containment isolation
valve OPERABILITY requirements or associated isolation time limits.
The proposed change does not involve a significant reduction in a
margin of safety because the ability to isolate containment in the
event of a release of radioactive material to the containment
atmosphere or pressurization of the containment will be maintained.
The margin of safety is defined by the established containment
isolation valve OPERABILITY requirements and associated isolation
time limits. The proposed change does not alter these operating
restrictions and the margin of safety which assures the ability to
isolate containment is not affected.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 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: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Project Director: George Dick, Acting.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: March 9, 1999.
Description of amendment request: The amendment request proposes
that reference to the Independent Safety Engineering Group be removed
from Technical Specification requirements, with supporting changes to
the Operations Quality Assurance Plan.
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 change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed amendment is a programmatic and administrative
change that
[[Page 17031]]
does not physically alter safety-related systems, nor does it affect
the way in which safety-related systems perform their functions. The
functions assigned to the Independent Safety Engineering Group are
addressed by other organizations. Because the design of the facility
and system operating parameters are not being changed, the proposed
amendment does not involve an increase in the probability or
consequences of any accident previously evaluated.
The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed amendment is a programmatic and administrative
change that does not physically alter safety-related systems, nor
does it affect the way in which safety-related systems perform their
functions. The functions assigned to the Independent Safety
Engineering Group are addressed by other organizations. Because the
design of the facility and system operating parameters are not being
changed, the proposed amendment does not create the possibility of a
new or different kind of accident previously evaluated.
The proposed change does not involve a significant reduction in
a margin of safety.
The proposed amendment is a programmatic and administrative
change that provides assurance that plant operations continue to be
conducted in a safe manner. The functions assigned to the
Independent Safety Engineering Group are addressed by other
organizations. As stated above the proposed amendment does not
physically alter safety-related systems, nor does it affect the way
in which safety-related systems perform their functions. Because the
design of the facility and system operating parameters are not being
changed, the proposed amendment does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 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: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Project Director: George F. Dick, Acting.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: March 15, 1999 (Supplement to October
29, 1998).
Description of amendment request: The proposed amendments were
submitted by application dated October 29, 1998, to relocate Technical
Specification (TS) 3/4.7.9 requirements for snubbers to the Technical
Requirements Manual. The Commission issued a Notice of Consideration of
Issuance of Amendments regarding its proposed no significant hazards
consideration determination that was published in the Federal Register
on December 16, 1998 (63 FR 69346).
Subsequently, by letter dated March 15, 1999, supplemental
information was submitted to include TS 6.10.3.l to be relocated to the
Technical Requirements Manual. This information is being noticed to
provide for public comment on the issue of no significant hazards
consideration.
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 or consequences of an accident previously evaluated?
The supplement to the amendment request relocates the record
keeping requirements of Technical Specification 6.10.3.l to the
Technical Requirements Manual. The change does not involve a
physical alteration of the plant (no new or different type of
equipment will be installed) or make changes in the methods
governing normal plant operation. The change will not impose
different requirements, and adequate control of information will be
maintained. This change will not alter assumptions made in the
safety analysis and licensing basis.
The Technical Requirements Manual is incorporated in the South
Texas Project Updated Final Safety Analysis Report and will be
maintained pursuant to 10 CFR 50.59. In addition, snubber
operability is addressed in existing surveillance procedures that
are also controlled by 10 CFR 50.59 and subject to the change
control provisions imposed by plant administrative procedures, which
endorse applicable regulations and standards.
Therefore, the supplement to the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The supplement to the amendment request relocates the record
keeping requirements of Technical Specification 6.10.3.l to the
Technical Requirements Manual. The change does not involve a
physical alteration of the plant (no new or different type of
equipment will be installed) or make changes in the methods
governing normal plant operation. The change will not impose
different requirements, and adequate control of information will be
maintained. This change will not alter assumptions made in the
safety analysis and licensing basis.
The Technical Requirements Manual is incorporated in the South
Texas Project Updated Final Safety Analysis Report and will be
maintained pursuant to 10 CFR 50.59. In addition, snubber
operability is addressed in existing surveillance procedures that
are also controlled by 10CFR50.59 and subject to the change control
provisions imposed by plant administrative procedures, which endorse
applicable regulations and standards.
Therefore, the change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The supplement to the amendment request relocates the record
keeping requirements of Technical Specification 6.10.3.l to the
Technical Requirements Manual. The relocated requirements remain the
same as the existing Technical Specifications. The change will not
reduce a margin of safety because it has no impact on any safety
analysis assumptions. Future changes to the relocated requirements
will be evaluated per the requirements of 10CFR50.59.
Therefore, the supplement will not result in 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 standards of 10 CFR 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: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Project Director: George F. Dick, Acting.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: February 15, 1999.
Description of amendment request: The proposed amendment would
revise requirements of Technical Specifications Section 6,
``Administrative Controls,'' related to (1) plant manager's
responsibilities, (2) plant staff titles and organization, (3) offsite
and onsite review committee (4) reportable events, and (5) actions
required in event of a safety limit violation.
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
[[Page 17032]]
consideration, which is presented below:
The proposed amendment will not change the intent of the TS or
decrease WPSC's management support or involvement in activities at
the Kewaunee Plant. Furthermore, it will not result in a decrease in
the engineering or technical support supplied by the plant staff or
the corporate support staff. The proposed changes are administrative
in nature. They primarily involve the relocation of existing
requirements to owner controlled documents; therefore, there are no
significant hazards associated with this change. As an
administrative change this will not result in a significant increase
in the probability of occurrence or consequences of an accident. As
an administrative change this will not create the possibility of a
new or different kind of accident from any previously analyzed. This
administrative change relocates existing requirements, and
therefore, will not involve a significant decrease in the margin of
safety.
In addition, the staff analyzed the proposed changes in accordance
with the provisions of 10 CFR 50.92. The proposed change will not:
1. Involve a significant increase in the probability or consequence
of an accident previously evaluated.
The analyses for the previously evaluated accidents are presented
in Chapter 14 of the Updated Safety Analysis Report. There are 19
postulated accidents addressed therein. The proposed amendment would
not affect the safety analysis assumptions or analytical models used
for any of these analyses. Also, the calculated dose consequences for
analyzed accidents would be unaffected. Therefore the proposed changes
do not involve a significant increase in the probability or consequence
of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed accident does not involve any physical change to the
design of the physicality, or operation of the facility outside the
bounds of the existing analyses. Thus, there is no possibility of
creating a new or different kind of accident.
3. Involve a significant reduction in the margin of safety.
The proposed changes do not involve any physical changes to any of
the fission product barriers or to the design or operation of any
safety systems. Also, no safety limits, limiting safety systems
settings, limiting conditions for operation or testing requirements
would be affected. Therefore, the proposed changes do not involve a
significant reduction in the margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Project Director: Cynthia A. Carpenter.
Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power
Station, Franklin County, Massachusetts
Date of amendment request: March 17, 1999.
Description of amendment request: Licensee submitted a License
Amendment request to delete administrative Technical Specification (TS)
requirements related to overtime restrictions. The licensee stated it
will provide appropriate constraints on excessive overtime in its
Administrative Procedures.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes are administrative in nature and simply
eliminate outdated requirements from the YNPS Technical
Specifications. As such the changes will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated. The administrative
nature of the changes will not affect safety-related systems or
components or their mode of operation and therefore, will not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Create the possibility of a new or different accident from
any previously evaluated. The proposed changes do not modify any
plant systems or components and, therefore, do not create the
possibility of a new or different accident from any previously
evaluated.
3. Involve a significant reduction in the margin of safety. The
changes are administrative in nature involving the deletion of
outdated requirements in the technical specifications; therefore,
there will be no reduction in the margin of safety.
Based on the considerations noted above, it is concluded that the
proposed changes will not endanger the public health and 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: Greenfield Community College,
1 College Drive, Greenfield, Massachusetts 01301.
Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One
International Place, Boston, Massachusetts 02110-2624.
NRC Project Director: Seymour H. Weiss.
Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power
Station, Franklin County, Massachusetts
Date of amendment request: March 17, 1999.
Description of amendment request: Licensee submitted a License
Amendment request to transfer Technical Specification Sections 6.7--
Procedures and Programs and 6.9--Record Retention to the Yankee
Decommissioning Quality Assurance Program (YDQAP).
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 changes are administrative in nature.
Administrative requirements in Sections 6.7 and 6.9 of the YNPS
Technical Specifications are to be transferred to the YDQAP which is
the current location of related administrative requirements. As such
the changes will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated. The administrative
nature of the changes will not affect safety-related systems or
components or their mode of operation and therefore, will not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Create the possibility of a new or different accident from
any previously evaluated. The proposed changes do not modify any
plant systems or components and, therefore, will not create the
possibility of a new or different accident from any previously
evaluated.
3. Involve a significant reduction in the margin of safety. The
changes are administrative in nature involving the relocation of
administrative requirements from one licensing document to another
licensing document currently containing related requirements;
therefore, there will be no significant reduction in the margin of
safety.
Based on the considerations noted above, it is concluded that the
proposed changes will not endanger the public health and safety.
The NRC staff has reviewed the licensee's analysis and, based on
this
[[Page 17033]]
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: Greenfield Community College,
1 College Drive, Greenfield, Massachusetts 01301.
Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One
International Place, Boston, Massachusetts 02110-2624.
NRC Project Director: Seymour H. Weiss.
Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power
Station, Franklin County, Massachusetts
Date of amendment request: March 17, 1999.
Description of amendment request: Licensee submitted a License
Amendment request to consolidate management positions and to transfer
Technical Specification review and audit functions to the Yankee
Decommissioning Quality Assurance Program (YDQAP).
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 changes are administrative in nature and reflect a
streamlining of the YAEC/YNPS management structure and procedures
consistent with the on-going requirement to complete the remaining
scope of YNPS decommissioning safely and efficiently. As such the
changes will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated. The administrative
nature of the changes will not affect safety-related systems or
components or their mode of operation and therefore, will not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Create the possibility of a new or different accident from
any previously evaluated. The proposed changes do not modify any
plant systems or components and, therefore, will not create the
possibility of a new or different accident from any previously
evaluated.
3. Involve a significant reduction in the margin of safety.
Elimination of the Manager of Operations position and the Plant
Superintendent position will not eliminate any of the
responsibilities or functions currently assigned to these positions.
These responsibilities or functions will be reassigned to an
appropriately qualified YAEC/YNPS manager, i.e., the Decommissioning
Manager. This change and replacement of the PORC and the NSARC
review and audit functions with an independent safety review and an
IRAC are consistent with the significant reduction in the scope and
the complexity of activities at YNPS as the facility moves into the
later stages of the decommissioning effort; therefore, there will be
no significant reduction in the margin of safety.
Based on the considerations noted above, it is concluded that the
proposed changes will not endanger the public health and 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: Greenfield Community College,
1 College Drive, Greenfield, Massachusetts 01301.
Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One
International Place, Boston, Massachusetts 02110-2624.
NRC Project Director: Seymour H. Weiss.
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.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: February 24, 1999.
Brief description of amendment: The amendment would revise
Technical Specification Table 3.3-1, ``Reactor Protective
Instrumentation,'' Action 2, for Arkansas Nuclear One, Unit No. 2. The
proposed change would add a footnote to Action 2 that would allow
startup and operation with the functional units associated with the
Channel ``D'' ex-core nuclear instrumentation to be maintained in the
bypassed or tripped condition following the restart from Refueling
Outage 2R13. This footnote is intended to support normal plant
operations until such time that the Channel ``D'' ex-core detector
assembly can be restored to an operable status.
Date of publication of individual notice in Federal Register: March
8, 1999 (64 FR 11067).
Expiration date of individual notice: April 7, 1999.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
[[Page 17034]]
CBS Corporation, Docket No. 50-22, Westinghouse Test Reactor, Waltz
Mill, Pennsylvania
Date of application for amendment: September 28, 1998 supplemented
on November 17, 1998.
Brief description of amendment: This amendment changes the license
to reflect the new legal name of the licensee for the Westinghouse Test
Reactor to CBS Corporation.
Date of issuance: March 25, 1999.
Effective Date: March 25, 1999.
Amendment No: 9.
Facility License No. TR-2: This amendment changes the license.
Date of initial notice in Federal Register: December 16, 1998, (63
FR 69334).
The Commission has issued a Safety Evaluation for this amendment
dated March 25, 1999.
No significant hazards consideration comments received: No.
Local Public Document: N/A.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units
1 and 2, Rock Island County, Illinois
Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2,
LaSalle County, Illinois
Date of application for amendments: December 17, 1998.
Brief description of amendments: The amendments revised the
respective facility Technical Specifications (TS) by adding a new
Limiting Condition for Operations that provided an administrative
enhancement by allowing testing required to return equipment to service
to be conducted under administrative controls.
Date of issuance: March 16, 1999.
Effective date: Immediately, to be implemented within 60 days.
Amendment Nos.: 172, 167; 184, 181; 132, 117.
Facility Operating License Nos. DPR-19, DPR-25, DPR-29, DPR-30,
NPF-11 and NPF-18.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 27, 1999 (64 FR
4153) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 16, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021; for LaSalle, the Jacobs Memorial Library, 815 North
Orlando Smith Avenue, Illinois Valley Community College, Oglesby,
Illinois 61348-9692.
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: August 14, 1998, as
supplemented on October 13 and December 23, 1998.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) to reflect the use of Siemens Power
Corporation (SPC) ATRIUM-9B fuel. Specifically, the amendments
incorporate the following into the TSs: (1) new methodologies that will
enhance operational flexibility and reduce the likelihood of future
plant derates; (2) administrative changes that adopt Improved Standard
Technical Specification (iSTS) language where appropriate; and (3)
changes to the Minimum Critical Power Ratio.
Date of issuance: March 16, 1999.
Effective date: Immediately, to be implemented prior to startup of
Cycle 9 for Unit 1 and prior to startup of Cycle 8 for Unit 2.
Amendment Nos.: 131, 116.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the TSs.
Date of initial notice in Federal Register: November 4, 1998 (63 FR
59588). The December 23, 1998, submittal 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
March 16, 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.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: August 14, 1998, as
supplemented by letters dated October 13, 1998, and December 23, 1998.
Brief description of amendments: The amendments changed the Quad
Cities Technical Specifications (TS) to reflect the use of Siemens
Power Corporation (SPC) ATRIUM-9B fuel. Specifically, the amendments
incorporate the following into the TS: (a) new methodologies that will
enhance operational flexibility and reduce the likelihood of future
plant derates; (b) administrative changes that eliminate the cycle-
specific implementation of ATRIUM-9B fuel and adopt Improved Standard
Technical Specification language where appropriate; and (c) changes to
the Minimum Critical Power Ratio (MCPR).
The amendment for Unit 1 also reflects the removal of Unit 1
specific pages incorporated into Unit 1 TS by Amendment No. 182 and are
no longer applicable. The August 14, 1998, application superseded an
August 29, 1997, application in its entirety (63 FR 2274).
Date of issuance: March 17, 1999.
Effective date: Immediately, to be implemented within 60 days.
Amendment Nos.: 185 & 182.
Facility Operating License Nos. DPR-29, DPR-30: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 9, 1998 (63
FR 48258) and November 4, 1998 (63 FR 59588). The October 13, 1998,
submittal changed a reference to a recently NRC-approved additive
constant uncertainty (ACU) generic methodology for ATRIUM-9B fuel (ANF-
1125 (P)(A), supplement 1, Appendix E) from Appendix D which provided
an interim value for ACU. This change was noticed on November 4, 1998
(63 FR 48258). The December 23, 1998, submittal 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 March 17, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: November 30, 1998.
Brief description of amendments: The amendments changed the
technical specifications (TSs) by decreasing the Allowed Outage Time
(AOT) from 67 days to 14 days for the Safe Shutdown Makeup Pump (SSMP).
Date of issuance: March 26, 1999.
Effective date: Immediately, to be implemented within 60 days.
Amendment Nos.: 186 & 183.
[[Page 17035]]
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 13, 1999 (64 FR
2246).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
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: September 30, 1998.
Brief description of amendments: The amendments increase the
maximum fuel rod internal pressure in the spent fuel pool from 1200
pounds per square inch gauge (psig) to 1300 psig by changing the
Updated Final Analysis Report (UFSAR) reference to the computer code
used to determine the fuel rod internal pressure (TACO3 computer code
would be added) in UFSAR Chapter 15. In addition, the amendments
justify not increasing the overall effective decontamination factor for
iodine as a consequence of a fuel handling accident and change the
terminology used in the UFSAR from ``fuel assembly gap gas pressure''
to ``fuel rod internal pressure.''
Date of Issuance: March 26, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1-301; Unit 2-301; Unit 3-301.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments authorized change(s) to the FSAR.
Date of initial notice in Federal Register: November 4, 1998 (63 FR
59590).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas.
Date of application for amendment: February 25, 1999.
Brief description of amendment: This amendment revises Technical
Specification (TS) Table 3.3-1, ``Reactor Protective Instrumentation,''
Action 2, for Arkansas Nuclear One, Unit No. 2 (ANO-2). This change
adds a footnote to Action 2 that allows startup and operation with the
functional units associated with the Channel ``D'' ex-core nuclear
instrumentation to be maintained in the bypassed or tripped condition
following the restart from Refueling Outage 2R13. This footnote is
intended to support normal plant operations until such time that the
Channel ``D'' ex-core detector assembly can be restored to an operable
status. This footnote will be in effect for a time period not to extend
beyond Mid-Cycle Outage 2P99, which is the next planned entry into cold
shutdown conditions for ANO-2. A Notice of Enforcement Discretion
(NOED) related to TS Table 3.3-1, Action 2, was issued verbally on
February 23, 1999. The NOED is documented in a letter dated February
25, 1999.
Date of issuance: March 23, 1999.
Effective date: As of the date of issuance.
Amendment No.: 202.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes (64 FR 11067 dated March 8, 1999). The notice
provided an opportunity to submit comments on the Commission's proposed
NSHC determination. No comments have been received. The notice also
provided for an opportunity to request a hearing by April 7, 1999, but
indicated that if the Commission makes a final NSHC determination, any
such hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, and final NSHC determination are contained in a
Safety Evaluation dated March 23, 1999.
Attorney for Licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington DC 20005-3502.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio.
Date of application for amendment: November 2, 1995, and as
supplemented by submittal dated January 7, 1999.
Brief description of amendment: This amendment revises technical
specification requirements for handling irradiated fuel in the Primary
Containment and the Fuel Handling Building, and selected specifications
associated with performing core alterations.
Date of issuance: March 11, 1999.
Effective date: March 11, 1999.
Amendment No.: 102.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 6, 1995 (60 FR
62497).
The supplemental information contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not change the scope of the original application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 11, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio.
Date of application for amendment: August 27, 1996, as supplemented
by submittals dated April 9, 1997, July 22, 1998, December 3, 1998, and
January 18, 1999.
Brief description of amendment: This amendment revised Technical
Specification 3.6.1.3, ``Primary Containment Isolation Valves
(PCIVs),'' and 3.6.1.9, ``Main Steam Isolation Valve (MSIV) Leakage
Control System (LCS).'' The amendment reflects implementation of the
revised accident source term in NUREG-1465, ``Accident Source Terms for
Light-Water Nuclear Power Plants'' and permits the licensee to
eliminate the MSIV LCS and increase the allowable leak rates of the
MSIVs.
Date of issuance: March 26, 1999:
Effective date: March 26, 1999.
Amendment No.: 103.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 7, 1998 (63 FR
53958).
The supplemental information contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 26, 1999.
No significant hazards consideration comments received: No.
[[Page 17036]]
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio.
Date of application for amendment: October 27, 1998.
Brief description of amendment: This amendment revised the minimum
critical power ratio (MCPR) safety limit contained in TS 2.1.1.2. In
addition, the amendment removes a note to TS 2.1.1.2 and a footnote to
TS 5.6.5.b that references MCPR safety limit values as cycle specific.
Date of issuance: March 26, 1999:
Effective date: March 26, 1999.
Amendment No.: 104.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 2, 1998 (63 FR
66603).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio.
Date of application for amendment: September 9, 1998, as
supplemented by submittals dated January 6, March 4, and March 18,
1999.
Brief description of amendment: This amendment revises the design
and licensing basis of containment isolation valves in the feedwater
system. The amendment revises (1) Surveillance Requirement 3.6.1.3.11
of Technical Specification (TS) 3.6.1.3, ``Primary Containment
Isolation Valves (PCIVs)'' to exclude the feedwater check valves from
the hydrostatic test program, (2) TS 5.5.2, ``Primary Coolant Sources
Outside Containment,'' to stipulate that water leakage past the
feedwater motor-operated containment isolation valves and the reactor
water cleanup system return to feedwater line is added to the program,
and (3) TS 5.5.12, ``Primary Containment Leakage Rate Testing
Program,'' to state that the feedwater check valves will be tested in
accordance with the Inservice Testing Program (TS 5.5.6).
Date of issuance: March 26, 1999.
Effective date: March 26, 1999.
Amendment No.: 105.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56262).
The supplemental information contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida.
Date of application for amendments: August 24, 1998.
Brief description of amendments: These amendments change the St.
Lucie Technical Specifications (TSs) by both removing obsolete license
conditions and revising the TSs. The amendments change the TSs to
modify the St. Lucie Unit 1 TSs to add components, not previously
described in the TSs, to the list of components that comprise an
operable control room emergency ventilation system, to modify the Unit
1 and Unit 2 TSs surveillance requirements to clarify component
operations, not previously described, that must be verified in response
to a containment sump recirculation actuation signal, to delete from
the facility operating license No. NPF-16 for Unit 2, license condition
2.C.19 to reflect the completion of the Unit 1 spent fuel pool re-rack
and delete license condition 2.I to reflect the resolution of
litigation and to modify license condition 2.B.5 to restore the
original syntax of the license condition and license condition 2.F to
update the references to current license conditions.
Date of Issuance: March 17, 1999.
Effective Date: These amendments shall be implemented within 30
days of receipt.
Amendment Nos.: 160 and 99.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the TSs.
Date of initial notice in Federal Register: September 23, 1998 (63
FR 50937).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 17, 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.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey.
Date of application for amendment: September 3, 1998.
Brief description of amendment: The amendment revises Technical
Specifications 3.4.A.10.e and 3.5.a.2.e to incorporate a Condensate
Storage Tank water level of greater than 35 feet.
Date of Issuance: March 17, 1999.
Effective date: March 17, 1999, to be implemented within 30 days
Amendment No.: 204.
Facility Operating License No. DPR-16. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6698).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated March 17, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit
1, DeWitt County, Illinois.
Date of application for amendment: January 20, 1999, as
supplemented February 4, 8, and 25, and March 5, 1999.
Brief description of amendment: The amendment changes the
undervoltage relay setpoints.
Date of issuance: March 26, 1999.
Effective date: March 26, 1999.
Amendment No.: 122.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 28, 1999 (64 FR
4474).
The four supplemental submittals provided additional information
and did not change the requested amendment or affect the proposed no
significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, IL 61727.
[[Page 17037]]
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of application for amendment: April 13, 1998, as supplemented
November 5, 1998.
Brief description of amendment: The proposed amendment would revise
the Appendix A Technical Specifications to base the Limiting Condition
for Operation for the fuel storage pool water level on a revised
analysis of the fuel handling accident and a new analysis for
radiological shielding during movement of irradiated fuel.
Date of issuance: March 16, 1999.
Effective date: March 16, 1999 (and shall be implemented no later
than 30 days).
Amendment No.: 162.
Facility Operating License No. DPR-36: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 20, 1998 (63 FR
27763). The November 5, 1998, submittal provided additional clarifying
information and did not change the initial proposed no significant
hazards determination and did not expand the scope of the original
application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 16, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578.
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point
Nuclear Station Unit No. 1, Oswego County, New York
Date of application for amendment: December 30, 1998.
Brief description of amendment: The amendment changes Technical
Specification (TS) Tables 3.6.14-2 and 4.6.14-2 regarding the noble gas
activity monitor channel operability requirement and daily sensor check
surveillance requirement to be consistent with the conditions specified
in TS 3.1.3.a for operability of the emergency cooling system. Also,
this amendment corrects a clerical error in TS 4.6.15.d.
Date of issuance: March 16, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 165.
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6699).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 16, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: November 19, 1998.
Brief description of amendment: This amendment changes surveillance
frequencies in Technical Specifications 4.8.4.4a and 4.8.4.5a to
require testing of the Electrical Protection Assemblies once every 6
months with the plant on-line rather than shut down.
Date of issuance: March 18, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 86.
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications
Date of initial notice in Federal Register: December 30, 1998 (63
FR 71970).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 18, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
North Atlantic Energy Service Corporation, et al., Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: May 20, 1998, as supplemented by letter
dated January 28, 1999.
Description of amendment request: Revise Technical Specifications
Table 3.3-4 and associated bases to depict a change to the refueling
water storage tank low-low level setpoint
Date of issuance: March 12, 1999.
Effective date: As of its date of issuance, to be implemented
within 60 days.
Amendment No.: 60.
Facility Operating License No. NPF-86. Amendment revised the
Technical Specifications
Date of initial notice in Federal Register: August 12, 1998 (63 FR
43205).
The supplemental letter 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 March 12, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: September 9, 1998, as
supplemented February 19 and 26, 1999.
Brief description of amendment: The amendment resolves several
previously identified technical specifications (TSs) compliance issues.
Specifically, the amendment: (1) changed TS definitions 1.24, ``Core
Operating Limits Report,'' 1.27, ``Engineering Safety Feature Response
Time,'' and 1.31, ``Radiological Effluent Monitoring and Offsite Dose
Calculation Manual (REMODCM)''; (2) changed TS 3.0.2, ``Limiting
Condition for Operation,'' by adding a new TS 3.0.6 to the Limiting
Condition for Operation TS section; (3) changed TS 4.0.5,
``Surveillance Requirements''; (4) changed the mode applicability of TS
3.2.3, ``Total Unrodded Integrated Radial Peaking--FrT'';
(5) changed TS 3.3.2.1, ``Engineered Safety Features Actuation System
Instrumentation,'' by modifying TS Table 4.3-2 Table Notation (1) which
it references; and (6) changed TS 3.4.1.1, ``Reactor Coolant System--
Reactor Coolant System Vents.'' The associated TS Bases sections were
also changed.
Date of issuance: March 11, 1999.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 230.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56251).
The supplemental letters provided clarifying information that did
not change the original proposed no significant hazards consideration
determination or expand the scope of the original Federal Register
notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 11, 1999.
[[Page 17038]]
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: July 17, 1998, as supplemented
November 10, 1998, and February 11, 1999.
Brief description of amendment: The amendment revises certain
diesel generator (DG) action statements and surveillance requirements
to improve overall DG reliability and availability.
Date of issuance: March 12, 1999.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 231.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 12, 1998 (63 FR
43207).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 12, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: November 25, 1997, as
supplemented September 25 and November 11, 1998, and January 28, 1999.
Brief description of amendment: The amendment revises the Technical
Specifications for the condensate storage tank (CST) low level suction
transfer setpoint for the high pressure coolant injection (HPCI) and
reactor core isolation cooling (RCIC) systems to allow removing one CST
from service for maintenance.
Date of issuance: March 19, 1999.
Effective date: March 19, 1999, with full implementation within 30
days.
Amendment No.: 105.
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 18, 1998 (63
FR 69344) The November 25, 1997, letter and September 25 and November
11, 1998, supplements were referenced in the original Federal Register
notice. The January 28, 1999, supplement provided an updated Technical
Specification page following the incorporation of Amendment 103, issued
December 23, 1998. This information was within the scope of the
original Federal Register notice and did not change the staff's initial
proposed no significant hazards considerations determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 19, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: November 25, 1998.
Brief description of amendments: The amendments revise Technical
Specifications 3.2 and Table 3.5-2B to allow limited inoperability of
boric acid storage tank level channels and transfer logic channels to
provide for required testing and maintenance of the associated
components.
Date of issuance: March 17, 1999.
Effective date: March 17, 1999, with full implementation within 30
days
Amendment Nos.: 143 and 134.
Facility Operating License Nos. DPR-42 and DPR-60. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 16, 1998 (63
FR 69345).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 19, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: July 30, 1997, as supplemented
by letter dated December 23, 1998.
Brief description of amendments: The amendments revise the combined
Technical Specifications (TS) for the Diablo Canyon Power Plant (DCPP)
Unit Nos. 1 and 2 by adding a Limiting Condition for Operation, trip
setpoints, and surveillance requirements for a residual heat removal
pump trip on refueling water storage tank level-low.
Date of issuance: March 26, 1999.
Effective date: March 26, 1999, to be implemented within 30 days
from the date of issuance.
Amendment Nos.: Unit 1--130; Unit 2--128.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 31, 1997 (62
FR 68312).
The December 31, 1997 supplemental letter provided additional
clarifying information, did not expand the scope of the application as
originally noted, and did not change the staff's proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: August 26, 1997, as
supplemented by letters dated October 14 and November 13, 1997, and
January 29, 1998.
Brief description of amendments: The amendments approve a
modification to the Diablo Canyon Power Plant (DCPP), Unit Nos. 1 and 2
auxiliary saltwater (ASW) system to bypass approximately 800 feet of
Unit 1 and 200 feet of Unit 2 Class 1 ASW pipe, a portion of which is
buried below sea level in the tidal zone outside the intake structure.
Date of issuance: March 26, 1999.
Effective date: March 26, 1999, and shall be implemented in the
next
[[Page 17039]]
periodic update to the FSAR Update in accordance with 10 CFR 50.71(e).
Amendment Nos.: Unit 1--131; Unit 2--129.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Final Safety Analysis Report Update.
Date of initial notice in Federal Register: September 16, 1997 (62
FR 48677).
The October 14 and November 13, 1997, and January 29, 1998,
supplemental letters provided additional clarifying information, 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 amendments is contained
in a Safety Evaluation dated March 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of application for amendments: September 25, 1996, as
supplemented on October 29, 1997, March 16, 1998, and February 8, 1999.
Brief description of amendments: The amendments revise the
Technical Specifications by revising the voltage and frequency
acceptance criteria and the start-timing methodology for the emergency
diesel generator surveillance testing.
Date of issuance: March 23, 1999.
Effective date: As of date of issuance, to be implemented within 60
days.
Amendment Nos: 218 and 200.
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 23, 1996 (61 FR
5039).
The October 29, 1997, March 16, 1998, and February 9, 1999, letters
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 March 23, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of application for amendments: September 17, 1998.
Brief description of amendments: The amendments revise Technical
Specification 3/4.8.2, ``Electrical Power Sources--Shutdown,'' for the
AC distribution system and the 125-volt and 28-volt DC distribution
systems. Specifically, the amendments change the Applicability and
Action Statements, if less than the complement of equipment and buses
are operable, to eliminate the need to establish containment integrity
and to add the action to suspend core alterations, positive reactivity
additions, and movement of irradiated fuel assemblies.
Date of issuance: March 24, 1999.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos.: 219 and 201.
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56257).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 24, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
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: April 6, 1995, as supplemented
on August 21, 1995. (TS 95-19).
Brief description of amendments: The amendments change the licenses
for Sequoyah Nuclear Plant, Units 1 and 2 by removing the license
conditions that reference the post-accident sampling system (PASS). The
PASS information has been placed in the Sequoyah Final Safety Analysis
Report (FSAR). This Change is consistent with NUREG-1431, ``Standard
Technical Specifications--Westinghouse Plants.''
Date of issuance: March 16, 1999.
Effective date: March 16, 1999.
Amendment Nos.: 243 and 233.
Facility Operating License Nos. DPR-77 and DPR-79: The amendments
revise the licenses.
Date of initial notice in Federal Register: April 26, 1995 (60 FR
20527). The August 21, 1995, letter provided clarifying information
that did not change the original no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 16, 1999.
No significant hazards consideration comments received: None.
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: April 23, 1998, as supplemented
on January 25, 1999.
Brief description of amendment: The amendment changes the existing
requirements for the Residual Heat Removal Service Water (RHRSW),
Station Service Water (SSW) and Alternate Cooling Tower Systems (ACS)
as identified in Technical Specifications (TSs) 4.5.C and 3/4.5.D.
Specifically, the changes are as follows:
(1) Specifications 3.5.D.3 and 4.5.D.3: This requirement is revised
to delete the existing allowance for 7 days of operation after both SSW
subsystems are made or found to be inoperable.
(2) Specification 4.5.C.1 and Specification 4.5.D.1: These
requirements have been revised to relocate testing information related
to pump flow and pressure testing characteristics for the RHRSW and SSW
Systems, respectively, to the Technical Requirements Manual.
(3) Specifications 3.5.D.1, 3.5.D.2, 3.5.D.3, 4.5.D.2, 4.5.D.3, and
associated Bases: All references to SSW ``subsystem'' have been
replaced by ``essential equipment cooling loop'' to more accurately
reflect the Vermont Yankee design and operation. In addition, certain
operability clarifications have been made to the Bases relative to
affected Specifications.
(4) Bases for Specification 3.5.D: The Bases have been revised to
omit statements that imply that the ACS could provide adequate heat
removal following a postulated accident. Other Bases additions have
been made that include certain operability clarifications relative to
affected Specifications.
Date of Issuance: March 11, 1999.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 169.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
[[Page 17040]]
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6713).
The January 25, 1999, supplement did not affect the original
proposed no significant hazards consideration.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated March 11, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, Vermont 05301.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: March 20, 1998, as supplemented by
letters dated May 28, June 30, August 28, September 4, November 20, and
December 8, 1998.
Brief description of amendment: The amendment revised the technical
specifications (TS) to support 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 of U-235. The amendment also
revised the TS to allow the storage of an additional 279 assemblies in
the cask loading pit.
Date of issuance: March 22, 1999.
Effective date: March 22, 1999, to be fully implemented no later
than December 31, 1999, except that the racks in the cask loading pit
may be installed at a future time after the completion of the next
refueling outage.
Amendment No.: 120.
Facility Operating License No. NPF-42: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 13, 1998 (63 FR
37601). The June 30, August 28, September 4, November 20, and December
8, 1998, supplemental letters provided additional clarifying
information, did not expand the scope of the application as originally
noticed, and did not change the staff's proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 22, 1999.
No significant hazards consideration comments received: No.
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.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: February 4, 1998, as supplemented by
letter dated October 20, 1998.
Brief description of amendment: The amendment revises the
requirements in Technical Specification Tables 3.3-3, 3.3-4 and 4.3-2
regarding the engineered safety features actuation system (ESFAS)
Functional Unit 6.f, and adds a note to Table 4.3-2 to clarify the
verification of time delays associated with ESFAS Functional Units 8.a
and 8.b.
Date of issuance: March 23, 1999.
Effective date: March 23, 1999, to be implemented within 30 days
from the date of issuance.
Amendment No.: 121.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 25, 1998 (63 FR
14491). The October 20, 1998, supplemental letter provided additional
clarifying information, 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 March 23, 1999.
No significant hazards consideration comments received: No.
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.
Dated at Rockville, Maryland, this 31st day of March 1999.
For the Nuclear Regulatory Commission.
Suzanne C. Black,
Acting Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 99-8503 Filed 4-6-99; 8:45 am]
BILLING CODE 7590-01-P