[Federal Register Volume 65, Number 46 (Wednesday, March 8, 2000)]
[Notices]
[Pages 12286-12299]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-5477]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from February 12, 2000, through February 25,
2000. The last biweekly notice was published on February 23, 2000 (65
FR 9000).
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
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By April 7, 2000, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC, and electronically from
the ADAMS Public Library component on the NRC Web site,
http://www.nrc.gov (the Electronic Reading Room). If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner
[[Page 12287]]
must provide sufficient information to show that a genuine dispute
exists with the applicant on a material issue of law or fact.
Contentions shall be limited to matters within the scope of the
amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner who fails
to file such a supplement which satisfies these requirements with
respect to at least one contention will not be permitted to participate
as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and electronically from the ADAMS Public
Library component on the NRC Web site,
http://www.nrc.gov (the Electronic Reading Room).
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of amendment request: November 30, 1999.
Description of amendment request: The proposed amendment revises
the test standard for laboratory testing of activated charcoal to tests
in accordance with the ASTM D3803-1989 standard in response to Generic
Letter 99-02, ``Laboratory Testing of Nuclear-Grade Activated
Charcoal.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed changes do not represent a significant increase
in the probability or consequences of an accident previously
evaluated.
The changes included in this request do not affect any accident
initiating events. No new accident initiators or new failure modes
are created. These changes will not result in any change to the
charcoal efficiency credited in the accident analyses for any of the
air treatment systems. The ability of each of the accident
mitigation air treatment systems to perform its function will not be
affected. System design flow requirements and filter/adsorber bank
bypass leakage requirements remain unchanged. Therefore, the
proposed changes will not adversely impact the capability of the
accident mitigation air treatment systems and could not represent a
significant increase in the probability or consequences of an
accident previously evaluated.
B. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This LCA [license change application] does not involve the
addition of any new hardware. The requested changes only affect
testing standards for the three air treatment systems used for
accident mitigation. Change[s] of a test standard for the air
treatment systems could not create a new accident scenario.
Therefore, these changes do not create the potential for any
accident different from those that have been evaluated.
C. These proposed changes do not involve a significant reduction
in a margin of safety.
The proposed T.S. changes will have no adverse affect on the
performance of the three accident mitigation Air Treatment Systems.
System design flow requirements and filter/adsorber bank bypass
leakage requirements remain unchanged. Use of the charcoal lab
testing protocol suggested by Generic Letter 99-02 will ensure that
the charcoal adsorber is better able to adsorb radioiodine generated
during postulated accidents. These changes do not result in a
degradation of safety related equipment, and therefore, do not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Edward J. Cullen, Jr., Esq., PECO Energy
Company, 2301 Market Street, S23-1, Philadelphia, PA 19103.
NRC Section Chief: Marsha Gamberoni, Acting.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendments request: November 19, 1999.
Description of amendments request: The proposed amendments would
revise Technical Specification (TS) 5.5.11.c, ``Ventilation Filter
Testing Program (VFTP),'' to change the testing requirements of the
engineered safety systems charcoal adsorbers.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Standard 1--Does the proposed change involve a significant increase
in the probability or consequences of an accident previously
evaluated?
No. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change to TS 5.5.11.c, initiates a laboratory
performance test of adsorber carbon (charcoal) that yields more
accurate results than what is currently required by TS. The proposed
change also deletes the specific reference to the ANSI [American
National Standards Institute] standard by which the adsorber carbon
sample is obtained. The proposed changes to test adsorber carbon to
a more current and improved ASTM [American Society for Testing and
Materials] standard and delete the ANSI standard by which the
adsorber carbon sample is obtained would not be plant accident
initiators as described in Chapter 6 or Chapter 15 of the PVNGS
[Palo Verde Nuclear Generating Station] UFSAR
[[Page 12288]]
[Updated Final Safety Analysis Report]. The changes would not
involve a significant increase in the probability of an accident
previously evaluated.
Carbon adsorption plays a direct role in mitigating the
consequences of a radiological event. Safety-related air-cleaning
units used in the ESF [engineering safety features] ventilation
systems of nuclear power plants reduce the potential onsite and
offsite consequences of a radiological accident by the adsorption of
radioiodine. The proposed amendment to change the laboratory
performance test for carbon will yield more conservative results
than what is currently required by TS. Hence, it will better ensure
that the adsorber carbon for TS systems used in the mitigation of an
accident remains above the assumed carbon decontamination efficiency
referenced in Chapter 6 and Chapter 15 of the UFSAR.
This proposed amendment does not alter, degrade, or prevent
actions described or assumed in an accident. It will not alter any
assumptions previously made in evaluating radiological consequences
or, affect any fission product barriers. It does not increase any
challenges to safety systems as well. Therefore, this proposed
amendment would not significantly increase the consequences of an
accident previously evaluated.
Standard 2--Does the proposed change create the possibility of a
new or different kind of accident from any accident previously
evaluated?
No. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change to TS 5.5.11.c, initiates a laboratory
performance test of adsorber carbon that yields more accurate
results than what is currently required by TS. The proposed changes
to test adsorber carbon to a more current and improved ASTM standard
and delete the specific reference to the ANSI standard by which the
adsorber carbon sample is obtained would not be plant accident
initiators as described in Chapter 6 or Chapter 15 of the PVNGS
UFSAR. The proposed amendment does not change the function of any
SSC [structure, system, and component]. TS nuclear air treatment
systems function to filter radiological releases during design basis
accidents. This change will provide greater assurance that this
function is provided. The revised TS required laboratory tests
utilize practices now in place, changing only the testing
parameters. The changes do not alter, degrade, or prevent actions
described or assumed in an accident described in the PVNGS UFSAR
from being performed. Therefore, the proposed amendment does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
Standard 3--Does the proposed change involve a significant reduction
in a margin of safety?
No. The proposed change does not involve a significant reduction
in a margin of safety.
The margin of safety, as defined in the PVNGS Technical
Specifications, is not reduced but is enhanced due to improved
testing. This change initiates a laboratory performance test on
adsorber carbon that yields more accurate results than what is
currently required by TS and deletes the specific reference to the
ANSI standard by which the adsorber carbon sample is obtained. The
proposed change to test adsorber carbon to a more current and
improved ASTM standard will ensure the carbon media's ability to
adsorb radioactive gases will remain above that credited in the
PVNGS' dose analysis for postulated accidents.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Section Chief: Stephen Dembek.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendments request: December 1, 1999.
Description of amendments request: The proposed amendments to the
operating licenses would delete or update outdated administrative
information and delete license conditions that are no longer applicable
or have been completed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Standard 1--Does the proposed change involve a significant increase
in the probability or consequences of an accident previously
evaluated?
No--The proposed administrative changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated. The proposed administrative Operating
License (OL) amendments would (1) delete or update operating license
references to outdated administrative information, (2) delete
license conditions that were complied with and are no longer
applicable to the current operating environment, and (3) delete
license conditions that were one-time requirements and have been
completed. Since these proposed changes are administrative and have
no [e]ffect on the current OL requirements, plant design, operation,
or maintenance, the proposed administrative changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Standard 2--Does the proposed change create the possibility of a
new or different kind of accident from any accident previously
evaluated?
No--The proposed administrative changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated. The proposed changes would have no [e]ffect on
the physical plant. Consequently, plant configuration and the
operational characteristics remain unchanged and the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Standard 3--Does the proposed change involve a significant reduction
in a margin of safety?
No--The proposed administrative changes do not involve a
significant reduction in a margin of safety. The proposed changes
are administrative and have no [e]ffect on the current OL
requirements, plant design, operation, or maintenance. No margin of
safety would be affected by the proposed administrative changes to
the PVNGS [Palo Verde Nuclear Generating Station] OLs since no
current operating requirements would be changed.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Section Chief: Stephen Dembek.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of amendments request: January 27, 2000.
Description of amendments request: The proposed amendment modifies
the conditions of containment closure during core alterations/fuel
handling and loss of shutdown cooling in Calvert Cliffs Units 1 and 2.
The reason for this proposed amendment is to improve personnel safety
and the progress of outages by allowing greater egress from and access
to the Containment during refueling outages. A new containment outage
door assembly will be installed on the outside of the equipment hatch
opening to provide quicker closure, improve safety when the door is
open, and allow more flexibility when staging material in the
Containment during an outage. Changes to the way the personnel air lock
and the containment
[[Page 12289]]
purge system are operated during maintenance activities on the Shutdown
Cooling System are also part of the proposed amendment.
The proposed amendment changes Technical Specifications (TSs) 3.9.3
and 3.9.4 to allow the new containment outage door to remain open
during core alterations and fuel handling, during maintenance and
testing activities on the Shutdown Cooling system, and to be used as an
alternate to the existing equipment hatch to close the equipment hatch
opening when containment closure is required. The proposed changes will
also allow the personnel air lock and the containment purge valves to
remain open during maintenance activities on the Shutdown Cooling
System. Baltimore Gas and Electric Company (BGE) also proposes to
revise TS 3.9.3 to indicate that four bolts is the minimum number
required to secure the equipment hatch for closure. In addition, BGE
proposes deleting the words ``when there is 23 feet of water above the
fuel'' from Limiting Condition for Operation 3.9.3.c.2 since this
requirement is already part of the applicability statement.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed changes will modify the conditions of containment
closure during core alterations/fuel handling and during
maintenance/testing activities on the Shutdown Cooling (SDC) System.
Specifically, the proposed changes will allow the new containment
outage door, the personnel air lock door, and the containment purge
valves to stay open during core alterations/fuel handling, and
during maintenance and testing activities on the SDC System. The
proposed change will also allow the new containment outage door to
be used as an alternate to the existing equipment hatch to close the
equipment hatch opening when closure is required. Additionally, the
proposed changes will change the wording of the Technical
Specifications to indicate that four bolts is the minimum number
required to secure the equipment hatch when it is used for
containment closure. The proposed changes also remove[ ] the water
level requirement from Limiting Condition for Operation 3.9.3 since
the water level requirement is part of the applicability statement
for this Technical Specification.
Closing the containment penetrations is considered to be a
mitigator of the radiological consequences of a fuel handling
incident and a loss of SDC, not an initiator. Therefore, allowing
the containment outage door, personnel air lock, and the containment
purge valves to be open during these outage activities does not
involve a significant increase in the probability of an accident
previously evaluated.
The consequence of a fuel handling incident is the release of
radioactivity from Containment. The potential offsite dose resulting
from a fuel handling incident has been evaluated. Based on a minimum
decay time of 100 hours prior to handling fuel (Technical Reference
Manual Section 15.9.1), the calculated offsite doses resulting from
a fuel handling incident are 14.06 rem to the thyroid, and 0.457 rem
to the whole body, with the personnel air lock door open. All
activity released from Containment over the length of the incident
is assumed to be unfiltered. The calculated doses resulting from a
fuel handling incident are less than 25% of the limits of 10 CFR
Part 100 (75 rem thyroid and 6 rem whole body). This analysis will
apply to the equipment hatch opening because the analysis assumes no
containment closure. The amount of radioactivity released is bounded
by the current analysis of record. Although natural air circulation
will cause some containment air to go out through any opening in a
fuel handling accident, there is no pressure produced to push the
radioactivity out of Containment. Therefore, having the containment
outage door open during core alterations and fuel handling does not
involve an increase in the consequences of an accident previously
evaluated. Additionally, if the equipment hatch is to be used,
specifying a minimum number of four bolts will allow the optional
use of more bolts, if desired.
The consequence of a loss of SDC is the potential for release of
radioactivity to the atmosphere outside Containment. Closing
containment penetrations is a mitigator of that consequence.
Administrative controls will be put in place to ensure that in an
emergency containment closure can be quickly achieved. The emergency
air lock will have at least one door closed when containment closure
is required by a SDC condition. The containment purge system
isolation valves are closed automatically on a containment high
radiation signal and can be shut by remote manual operation. The
maximum calculated pressure that can develop in the Containment for
the limiting loss of SDC case is 12 psig. All required penetration
closure devices can withstand that pressure. Therefore, allowing the
personnel air lock doors, the containment outage door, and the purge
isolation valves to remain open does not involve a significant
increase in the consequences of a loss of SDC.
Therefore, the proposed Technical Specificaton changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
This requirement change does not involve a significant change in
the operation of the plant and no new accident initiation mechanism
is created by the modification. Closing containment penetrations is
considered to be a mitigator of the radiological consequences of any
accident in the Containment, not an initiator. The equipment hatch
opening, the personnel air lock, and the purge supply and exhaust
are currently opened and closed during the course of an outage. The
proposed changes allows them to remain open during a period when
they are currently required to be closed. The closure function of
the equipment hatch opening in Modes 5 and 6 will be performed by a
hinged containment outage door; thus, closing the equipment hatch
opening will be easier and will require fewer people and less time.
The operation of the containment outage door is not a significantly
different method of operation from that of other dogged doors at
Calvert Cliffs. Using the containment outage door to close the
equipment hatch opening instead of the equipment hatch also
mitigates the consequences of the incident and does not initiate an
accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in the margin of
safety.
The margin of safety for containment closure during core
alternation/fuel handling is based on the amount of offsite dose
resulting from a fuel handling incident and the safety of personnel
in the Containment at the time of the incident. An offsite dose
calculation previously approved by the NRC for a fuel handling
incident is 14.06 rem to the thyroid, and 0.457 rem to the whole
body, with no containment closure established, and any activity
released from the Containment assumed to be unfiltered. These
calculated doses are less than 25% of the limits of 10 CFR Part 100.
The analysis will apply to the containment outage door because the
analysis assumes no containment closure. Emergency personnel egress
from Containment will be through the open door, which is an
improvement in personnel safety because this exit is not currently
available. Additionally, trained personnel will be available to
close the door and contain any radiation released inside Containment
as a result of a fuel handling incident. Leaving the containment
outage door open during core alterations and fuel handling will not
allow more than the calcutated amount of radionuclides to escape
from Containment; shutting the door following a fuel handling
incident will increase the margin of safety by keeping the actual
offsite dose lower than the calculated dose.
Therefore, allowing the containment outage door to be open
during fuel handling would not involve a significant reduction in
the margin of safety.
The margin of safety for containment closure in the case of loss
of SDC is twofold: (1) the time required to close the Containment to
prevent a radioactive release to the atmosphere outside Containment
if SDC should be lost; and (2) the ability to retain the pressure
generated by boiling of reactor coolant as a result of a loss of
SDC.
Currently, all containment penetrations are required to be
closed prior to taking the SDC
[[Page 12290]]
System out-of-service for maintenance, or within four hours if SDC
is lost. The radiological consequences of a loss of SDC incident do
not occur immediately on loss of SDC. The containment purge
isolation valves close rapidly on a high radiation signal or are
closed by remote manual operation. The containment outage door and
the personnel air lock doors are designed to be closed rapidly by
site personnel. Other containment penetrations that could release
radiation to the environment outside the Containment will be
required to be closed. The maximum calculated pressure that can
develop in the containment as a result of a loss of SDC is 12 psig.
The purge isolation valves, the personnel air lock doors, and the
containment outage door are all designed to meet this pressure
retaining requirement. The proposed changes do not increase the
possibility of a release of radiation following a loss of SDC
incident.
Therefore, the ability to provide containment closure is
maintained and the margin of safety is not significantly reduced by
this proposed activity.
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
amendments request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Marsha Gamberoni, Acting.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: January 11, 2000.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) to increase allowable out-of-
service times (AOTs) and surveillance test intervals (STIs) for
selected actuation instrumentation. The proposed amendments implement
AOT/STI changes based on Topical Reports by General Electric Company
and the Boiling Water Reactor Owners' Group that have previously been
reviewed and approved by the NRC.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Does the change involve a significant increase in the probability
or consequences of an accident previously evaluated?
The proposed TS changes increase the Allowable Outage Times and
Surveillance Test Intervals (AOT/STI) for actuation instrumentation
based on analyses developed and approved by the Nuclear Regulatory
Commission (NRC). TS requirements that govern operability or routine
testing of plant instruments are not assumed to be initiators of any
analyzed event because these instruments are intended to prevent,
detect, or mitigate accidents. Therefore, these changes will not
involve an increase in the probability of occurrence of an accident
previously evaluated. Additionally, these changes will not increase the
consequences of an accident previously evaluated because the proposed
changes do not involve any physical changes to plant systems,
structures or components (SSCs), or the manner in which these SSCs are
operated. These changes will not alter the operation of equipment
assumed to be available for the mitigation of accidents or transients
by the plant safety analysis or licensing basis. As justified and
approved in the AOT/STI licensing topical reports, the proposed changes
establish or maintain adequate assurance that components are operable
when necessary for the prevention or mitigation of accidents or
transients and that plant variables are maintained within limits
necessary to satisfy the assumptions for initial conditions in the
safety analyses. The proposed changes establish or modify time limits
allowable for operation with inoperable instrument channels based on
analyses which have been approved by the NRC. Furthermore, there will
be no change in the types or significant increase in the amounts of any
effluents released offsite. For these reasons, the proposed changes do
not involve a significant increase in the probability or consequences
of an accident previously evaluated.
Does the change create the possibility of a new or different kind
of accident from any accident previously evaluated?
The proposed changes do not involve any physical changes to SSCs,
or the manner in which these SSCs function. Therefore, these changes
will not create the possibility of a new or different kind of accident
from any accident previously evaluated. The changes in methods
governing normal plant operation are consistent with the current safety
analysis assumptions. Therefore, these changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
The proposed changes increase the AOTs and STIs for actuation
instrumentation based on generic analyses completed by the Boiling
Water Reactors Owners' Group (BWROG). The NRC has reviewed and approved
the generic studies and has concurred with the BWROG that the proposed
changes do not significantly affect the probability of failure or
availability of the affected instrumentation systems. The analysis
determined that there is no significant change in the availability and/
or reliability of instrumentation as a result of the proposed changes
in AOTs and STIs.
Furthermore, the change to increase the frequency of the reactor
protection system scram contactor testing has been shown to improve
plant safety. ComEd has determined these studies are applicable to
Dresden Nuclear Power Station, Units 2 and 3. The proposed changes to
AOTs provide realistic times to complete required testing and
maintenance actions without increasing the overall instrument failure
frequency. Likewise, the extended STIs do not result in significant
changes in the probability of instrument failure. Furthermore, the
proposed changes will reduce the probability of test-induced plant
transients and equipment failures. Therefore, it is concluded that 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 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767
NRC Section Chief: Anthony J. Mendiola.
Entergy Nuclear Generation Company, Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: November 22, 1999.
Description of amendment request: The proposed amendment would
approve the use of new values for post-accident containment pressure in
Pilgrim's net positive suction head (NPSH) analyses performed for the
emergency core cooling system (ECCS) pumps.
Basis for proposed no significant hazards consideration
determination:
[[Page 12291]]
As required by 10 CFR 50.91(a), the licensee has provided its analysis
of the issue of no significant hazards consideration, which is
presented below:
(1) Will crediting the proposed post-LOCA [loss of coolant
accident] containment pressure in ECCS analysis involve a
significant increase in the probability or consequences of an
accident previously evaluated?
Chapter 14 of the FSAR [final safety analysis report] contains
evaluations of the worst postulated accidents that the Pilgrim plant
was evaluated for, which include the refueling accident, the main
steam line break outside primary containment, the recirculation line
break inside primary containment, and the control rod drop accident.
No increase in the probability of the evaluated accidents will
result from crediting the proposed containment pressure because
post-LOCA containment pressure does not represent an accident
initiator but, rather, is an expected condition that will inherently
exist in the containment after the pipe break inside containment.
The worst radiological consequences for the Pilgrim plant are
associated with the design basis LOCA which is the double guillotine
failure of the recirculation system piping. The radiological
analysis of this event contained in FSAR Chapter 14 uses a TID-14844
source term and assumes a 1.5% per day leakage from the containment,
which is greater than the maximum leakage allowed by the Technical
Specifications. The results of this analysis are presented in Table
14.5-2 of the FSAR and indicate substantial margin when compared to
10 CFR Part 100 limits.
The radiological consequences of the design basis accident are
not increased by taking credit for the post-LOCA suppression pool
overpressure. Assuming containment integrity exists, the mechanism
for increasing the consequences of the accident would be an
increased leakage rate caused by an increase of the average
differential pressure between primary and secondary containment
during the accident response. However, the NPSH analysis performed
for Pilgrim that includes post-LOCA containment pressure does not
assume or require that the differential pressure between primary and
secondary containment be maintained above the lower bounding minimum
that exists due to thermal equilibrium conditions between the
containment atmosphere and the suppression pool. Specifically, the
containment pressure included in the ECCS pump NPSH analysis is
inherently provided by the increase in wetwell vapor pressure and
air/nitrogen partial pressure that exists due to equilibrium with
increasing pool temperature with an accounting for containment
initial conditions and leakage.
Inclusion of the post-LOCA containment pressure in the
calculation of NPSH does not require that a higher containment
pressure than would otherwise occur be purposely maintained, no
requirement is incurred to delay operating containment heat removal
equipment at the highest rate possible, no requirement is incurred
to deliberately continue any condition of high containment pressure
to maintain adequate NPSH, and no requirement is incurred for the
purposeful addition of air/nitrogen into the containment to increase
the available pressure.
The higher debris head losses that required the new NPSH
evaluation are based on an updated analysis of LOCA-generated
debris. The new debris analysis was performed in response to NRC
Bulletin 96-03 using the guidance given in Regulatory Guide 1.82,
Revision 2. The NRC guidance is used to ensure sufficient NPSH
margin exists to accommodate the debris resulting from a LOCA. Using
the proposed containment pressure limits included in this submittal,
it is shown there is sufficient NPSH margin at all times following
the bounding design basis accident.
Based on these reasons, the probability of accidents previously
evaluated is not increased and the consequences of the design basis
accident are not increased.
(2) Will crediting the proposed post-LOCA containment pressure
create the possibility for new or different kinds of accidents?
As stated above, Chapter 14 of the FSAR contains the worst
postulated accidents that the Pilgrim plant was evaluated for, which
include the refueling accident, the main steam line break outside
primary containment, the recirculation line break inside primary
containment, and the control rod drop accident. New or different
types of accidents are not created by including the containment
pressure in NPSH analyses because post-LOCA containment pressure is
an expected condition that will exist in the containment after the
pipe break inside containment. The pressure included in the NPSH
analysis is the minimum pressure that will exist due to thermal
equilibrium conditions and must be considered as part of any
accident analysis regardless of whether it is used in the evaluation
of NPSH.
(3) Will crediting the proposed new limits for post-LOCA
containment pressure in ECCS NPSH analyses involve a significant
reduction in a margin of safety?
The integrity of the primary containment and the operation of
the ECCS systems in combination limit the off-site doses to values
less than those suggested in 10 CFR 100 in the event of a break in
the primary system piping. In order for the ECCS pumps to meet their
performance requirements, the NPSH available to the pumps throughout
the accident response must meet their specific NPSH requirements.
Excess NPSH margin will not improve the performance of the ECCS
pumps because NPSH available must only meet NPSH requirements for
the pump to operate on its pump curve and meet design expectations.
Including the proposed post-LOCA containment pressure in NPSH
analyses increases the NPSH available to the ECCS pumps, but the
methodology used includes only that pressure that will inherently
exist due to thermal equilibrium between the containment atmosphere
and the suppression pool because of the primary containment
enclosure with an accounting for leakage. Post-accident containment
pressure calculated in such a manner represents a conservative lower
bound for the pressure that will be available. Therefore, it is
expected the actual NPSH margin will exceed that calculated by these
methods. The proposed pressure limits are enveloped at all times by
the containment pressure calculated using the thermal equilibrium
methodology. These methods for calculating NPSH available and NPSH
margin were previously reviewed by the NRC for License Amendment
173.
The new debris analysis referenced in this submittal was done in
accordance with Regulatory Guide 1.82, Revision 2. The LOCA debris
analysis is considered conservative and bounding for all postulated
accidents and transients. It is shown that, within the proposed
containment pressure limits, there is sufficient NPSH margin at all
times following the design basis accident to accommodate the debris
head loss without affecting RHR [residual heat removal] or core
spray pump performance.
Based on the above discussion, credit for the updated values of
containment pressure in ECCS NPSH analyses does 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: W. S. Stowe, Esquire, Entergy Nuclear
Generation Company, 800 Boylston Street, 36th Floor, Boston,
Massachusetts 02199.
NRC Section Chief: James W. Clifford.
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Units 1 and 2 (ANO-1&2), Pope County, Arkansas; Entergy
Operations, Inc., System Energy Resources, Inc., South Mississippi
Electric Power Association, and Entergy Mississippi, Inc., Docket No.
50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County,
Mississippi; Entergy Gulf States, Inc., and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish,
Louisiana; and Entergy Operations Inc., Docket No. 50-382, Waterford
Steam Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: November 23, 1999.
Description of amendment request: The proposed amendments would
incorporate the use of American Society for Testing and Materials
(ASTM) D3803-1989, ``Standard Test Method for Nuclear-Grade Activated
Carbon,'' into each facility's Technical Specifications (TS). Entergy
Operations, Inc. (the licensee) is submitting these proposed amendments
as a complete response to NRC Generic Letter (GL) 99-02, ``Laboratory
Testing of Nuclear-Grade Activated Charcoal.''
[[Page 12292]]
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:
I. The proposed change does not signficantly increase the
probability or consequences of an accident previously evaluated.
The proposed changes do not result from a physical change to the
facilities or impact plant operations. Neither do they impact the
response of the facilities to an accident.
American Society of Testing and Materials (ASTM) D3803-1989,
``Standard Test Method for Nuclear-Grade Activated Carbon,''
reflects the most up-to-date method for accurately testing the
efficiency of activated charcoal contained in engineered safety
features (ESF) system adsorbers. Establishing ASTM D3803-1989 as the
required method for laboratory testing of activated charcoal
represents an upgrade from the current TS requirements. Using ASTM
D3803-1989 methodology ensures the tested charcoal will perform in a
manner consistent with the facility's licensing basis.
The proposed acceptance criterion values for charcoal efficiency
were calculated using the equation specified in GL 99-02. As
documented in GL 99-02, the NRC [Nuclear Regulatory Commission]
found this equation acceptable for determining charcoal efficiency
when using ASTM D3803-1989 as the test method.
Based on the above discussion, the proposed changes do not
significantly increase the probability or consequences of an
accident previously evaluated.
II. The proposed change does not create the possiblity of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not result from a physical change to the
facilities or impact plant operations.
Establishing ASTM D3803-1989 as the method for performing
laboratory testing of nuclear-grade activated charcoal does not
involve a physical alteration to the facility or impact plant
operations. Using ASTM D3803-1989 methodology ensures the tested
charcoal will perform in a manner consistent with the facility's
licensing basis.
The proposed acceptance criterion values for charcoal efficiency
were calculated using the equation specified in GL 99-02. As
documented in GL 99-02, the NRC found this equation acceptable for
determining charcoal efficiency when using ASTM D3803-1989 as the
test method.
Based on the above discussion, the proposed changes do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
III. The proposed change does not involve a significant
reduction in a margin of safety.
The proposed changes do not result from a physical change to the
facilities or impact plant operations. Neither do they impact the
response of the facilities to an accident.
Safety-related air-cleaning units used in the ESF ventilation
systems of nuclear power plants reduce the potential onsite and
offsite consequences of a radiological accident by adsorbing
radioiodine. To ensure the charcoal adsorbers used in these systems
perform in a manner that is consistent with the facility's licensing
basis, facility TS contain requirements to periodically test (in a
laboratory) samples of charcoal taken from the air-cleaning units.
ASTM D3803-1989 reflects the most up-to-date method for
accurately testing the efficiency of activated charcoal contained in
ESF system adsorbers. Establishing ASTM D3803-1989 as the required
method for laboratory testing of activated charcoal represents an
upgrade from the current TS requirements and maintains the margin of
safety by ensuring the tested charcoal performs in a manner
consistent with the facility's licensing basis.
The proposed acceptance criterion values for charcoal efficiency
were calculated using the equation specified in GL 99-02. As
documented in GL 99-02, the NRC found this equation acceptable for
determining charcoal efficiency when using ASTM D3803-1989 as the
test method.
Based on the above discussion, the proposed changes do not
involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502; and Mark
Wetterhahn, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC
20005.
NRC Section Chief: Robert A. Gramm.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of amendment request: November 23, 1999.
Description of amendment request: The proposed amendments would
make the following changes to the Beaver Valley Power Station, Unit
Nos. 1 and 2 (BVPS-1 and BVPS-2) Technical Specifications (TSs): (1)
For BVPS-1, surveillance requirement (SR) 4.8.1.1.2.b.3.b would be
revised to reflect a narrower required diesel generator (DG) frequency
band; an associated footnote would be deleted; associated Bases would
be revised to reflect these TS changes. (2) For BVPS-2, SR 4.8.1.1.2.f
would be revised to clarify that the DGs are only required to achieve a
minimum frequency and voltage within the first 10 seconds of the
related test, and that the stated voltage and frequency bands are
requirements for steady state operation of the DGs; a footnote is also
added to this SR. (3) Page formats are revised as needed to permit the
addition or deletion of text.
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?
For the Beaver Valley Power Station (BVPS) Unit No. 1 only, the
proposed amendment will revise surveillance requirement (SR)
4.8.1.1.2.b.3.b.
Specifically, the required diesel generator (DG) frequency band
specified in SR 4.8.1.1.2.b.3.b will be reduced. In addition,
Footnote (6) pertaining to the DG frequency limits and associated
Bases wording will be deleted.
For BVPS Unit No. 2 only, SR 4.8.1.1.2.f will be revised to
clarify that the diesel generators are only required to achieve a
minimum voltage and frequency in 10 seconds. The DGs are
then required to obtain voltages and frequencies within the required
bands during steady state operation. A new Footnote (8) will be
added which modifies the stated voltage and frequency values in the
proposed SR 4.8.1.1.2.f.1. This footnote will require the voltage
and frequency values be appropriately increased to account for
measurement uncertainties.
Page format will be revised as necessary to permit incorporation
and deletion of text. These format changes include the addition or
deletion of Technical Specification pages as required.
The DGs are used to support mitigation of the consequences of a
design basis accident (DBA); however, they are not considered the
initiator of any previously analyzed DBA described in the Updated
Final Safety Analysis Report (UFSAR). The proposed amendment does
not impact any of the offsite AC distribution system; therefore, the
probability of a loss of offsite power event is not increased.
Therefore, the proposed amendment does not involve a significant
increase in the probability of an accident previously evaluated.
The proposed reduction in the DG output frequency limits (for
BVPS Unit No. 1 only) will continue to protect engineered safety
feature (ESF) pumps from runout conditions and ESF pump motors from
operating in an unanalyzed condition. The revised frequency limits
have no adverse effect on the diesel generator operability. The
revised DG output frequency limits do not increase the consequences
of a design basis accident; this proposed change ensures that
equipment will perform its intended function. This change is
intended to prevent the DG from being loaded beyond analyzed loading
limits and protect ESF equipment. The revised surveillance
requirements being applied to operating limits will provide greater
[[Page 12293]]
assurance that increased performance requirements are not imposed on
ESF equipment.
The proposed deletion of Footnote (6) (for BVPS Unit No. 1 only)
removes the ability to evaluate the DG frequency response. The
proposed wording is more restrictive in that the DG frequency
response will be required to be demonstrated regardless of the
amount of DG loading. The ability of the DGs to maintain the
required output frequency as required to meet accident analysis
assumptions will continue to be demonstrated on a periodic basis.
The proposed deletion of the Bases wording pertaining to Footnote
(6) is administrative in nature and does not affect plant safety.
This change removes guidance information on how to conduct the
engineering evaluation that will no longer be applicable following
DG governor modifications.
The proposed revision to SR 4.8.1.1.2.f (for BVPS Unit No. 2
only) will continue to require that both DGs start simultaneously to
confirm that there is not a cross-tie that could render both DGs
incapable of performing their required functions. The proposed
revision to SR 4.8.1.1.2.f will continue to require that each DG
obtain the minimum conditions to accept load in the time frame
assumed in the accident analysis. In addition, the proposed wording
of SR 4.8.1.1.2.f will continue to require that each DG obtain the
required steady state voltage and frequency values.
The proposed addition of Footnote (8) (for BVPS Unit No. 2 only)
is administrative in nature and does not affect plant safety. The
proposed footnote provides information that the values for voltage
and frequency need to be increased to account for measurement
uncertainties.
The revision to page format as necessary to permit incorporation
and deletion of text is editorial in nature and does not affect
plant safety.
Therefore, the proposed amendment does not involve a significant
increase in the 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 proposed revisions do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed revisions have no adverse impact on the DBAs previously
evaluated in the UFSAR. The proposed revisions will continue to
assure that the DGs are available and fully operable to perform
their intended safety function of providing sufficient electrical
power to ESF equipment following a DBA and a loss of offsite power.
New failure modes are not introduced as a result of the proposed
revisions to the DG surveillance requirements. The proposed revision
to the required DG frequency range will continue to prevent ESF
motors and pumps from being subjected to overfrequency conditions
which could reduce the life of the equipment. The proposed changes
do not affect the probability of malfunction of a DG or its
connected emergency AC power system.
Therefore, the proposed amendment 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 margin of safety is not significantly reduced as a result of
the proposed revisions. The margin of safety depends on the
maintenance of specific operating parameters within design limits.
The BVPS Unit No. 1 DG reliability and performance during a loss
of offsite power and a DBA are enhanced by the proposed revision to
SR 4.8.1.1.2.b.3.b. This proposed revision (for BVPS Unit No. 1
only) will ensure that the maximum calculated DG loading does not
exceed the UFSAR limit of 2745 kW. The proposed revision to SR
4.8.1.1.2.b.3.b for DG operating frequency limits continues to
protect ESF equipment from overfrequency conditions. ESF equipment
will continue to function, as assumed in the safety analysis, to
ensure that fuel, reactor coolant system and containment design
limits are not exceeded.
The proposed revision to SR 4.8.1.1.2.f (for BVPS Unit No. 2
only) will continue to require that both DGs start simultaneously to
confirm that there is not a cross-tie that could render both DGs
incapable of performing their required functions. The proposed
revision to SR 4.8.1.1.2.f will continue to require that each DG
obtain the minimum conditions to accept load in the time frame
assumed in the accident analysis. In addition, the proposed wording
of SR 4.8.1.1.2.f will continue to require that each DG obtain the
required steady state voltage and frequency values.
The remaining changes are either administrative or editorial in
nature and do not affect plant safety.
Therefore, the proposed amendment does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Marsha Gamberoni.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of amendment request: April 15, 1999, as supplemented on
December 22, 1999.
Description of amendment request: The proposed amendment would
change the Technical Specification as described below:
Page 1.0-3 Clarification would be added to the definition of
Secondary Containment Integrity.
Page 1.0-4 The definition of facility description and safety
analysis report (FDSAR) would be expanded.
Page 2.3-3 The Bases section would be separated from this page
which is the last page of the specification.
Page 2.3-4 Two paragraphs, which should have been deleted in an
earlier revision, would be deleted and subsequent pagination would
be affected. Two paragraphs would be moved from the end of the bases
to that location. An unrelated wording change would also be made.
Page 2.3-7 In addition to pagination, a sentence would be added
about the relays involved in undervoltage situations.
Page 3.4-1 The phrase ``(see Note below)'' would be deleted as
unnecessary and two lines from the top of page 3.4-2 would be
included as ``c.''
Page 3.4-2 Two lines would be moved to the prior page and
designated ``c.''
Page 3.5-7 LCO statement of 36 hours would be deleted from
Specification 3.5.B.6.a.3 because it is inconsistent with
Specification 3.5.B.7.
Page 3.5-9 A bases statement would be added about administrative
control over non-automatic primary containment isolation valves.
Page 3.5-11 A bases statement would be added about the use of the
trunion room door.
Page 3.7-1 The phrase ``shutdown position'' would be corrected to
``shutdown condition.''
Page 3.17-1 The phrase ``the control room HVAC system'' would be
corrected to ``one control room HVAC system.''
Page 4.5-13 The word ``off'' would be changed to ``on.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change are relatively minor in nature and are proposed
to enhance clarity and understanding. None of the changes have any
impact on safety and there is no change to an operating parameter of
any system, component or structure. Accordingly, the proposed changes
do not affect any accident precursors. Therefore, the probability of an
accident previously evaluated is not increased. The proposed TS change
will assure the ability of systems to perform their intended function.
Therefore, the proposed changes will not involve a significant increase
in the consequences of an accident previously evaluated. Therefore, the
probability of occurrence
[[Page 12294]]
or the consequences of an accident previously evaluated in the Safety
Analysis Report (SAR) will not increase as a result of these changes.
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 are relatively minor in nature and are
proposed to enhance clarity and understanding. None of the changes have
any impact on safety and there is no change to an operating parameter
of any system, component or structure. The proposed changes do not
involve placing systems in new configurations or operating systems in a
different manner that could result in a new or different kind of
accident. Therefore, the proposed activity does not create the
possibility for a new or different kind of accident from any previously
identified in the SAR.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The proposed changes do not involve a significant reduction in the
margin of safety. The changes are primarily administrative and are
proposed to enhance clarity and understanding. They do not modify an
operating parameter of any system, component or structure. They do not
adversely affect the performance characteristics of systems nor do they
affect the ability of systems to perform their intended function.
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.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: M. Gamberoni, Acting.
PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating
Station (LGS), Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: December 15, 1999.
Description of amendment request: The proposed changes will revise
the LGS Technical Specifications (TSs) to remove TS Table 3.6.3-1,
``Primary Containment Isolation Valves,'' and references to the table,
from the TSs and relocate the information from the TS table to the
Technical Requirements Manual, a licensee-controlled document.
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 probability of occurrence of a previously evaluated accident
is not increased because containment isolation is not an accident
initiator and the proposed changes do not impact any accident
initiating conditions. The consequences of an accident previously
evaluated are not increased because the proposed changes do not
impact the ability of containment to restrict the release of any
fission product radioactivity to the environment. The proposed
change to remove the primary containment isolation valve table from
TS and relocate the information to an administratively controlled
document, and to revise the wording in TS to reflect this change,
will have no impact on any safety related structures, systems or
components. The Technical Specification requirements for the primary
containment isolation valves will not be changed. In addition, the
details of the table are not being changed, only relocated to a
different controlling document. The proposed changes simplify the
Technical Specifications, meet the regulatory requirements for
control of containment isolation, and are consistent with the
guidance provided in Generic Letter 91-08. Therefore, the proposed
changes do not involve a significant 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 proposed changes are administrative in nature and do not
result in physical alterations or changes in the method by which any
safety related system performs its intended function(s). The
proposed changes do not impact any safety analysis assumptions. The
proposed changes do not create any new accident initiators or
involve an activity that could be an initiator of an accident of a
different type. Therefore, the proposed 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 a margin of safety.
The proposed change to remove the primary containment isolation
valve table from TS and relocate the information to an
administratively controlled document, and to revise the wording in
TS to reflect this change, do not alter the Technical Specifications
requirements for containment integrity and containment isolation and
will not adversely affect the containment isolation capability. The
licensee controlled document will be maintained under the
requirements of TS Administrative Controls Section 6.0 and the
provisions of 10CFR50.59. In addition, the proposed changes do not
impact any safety analysis assumptions. Therefore, the proposed
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101. NRC. Section Chief: James W. Clifford.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of amendment request: February 9, 2000.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Limiting Condition for Operation
(LCO) 3.8.2.1. The proposed change would add two new Action Statements
for operating conditions where a Class-1E battery's electrolyte
temperature is below the minimum limit specified in TS Surveillance
Requirement 4.8.2.1.b.3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed TS change does not involve any physical changes to
plant structures, systems or components (SSC). The Class-1E
batteries will continue to function as designed. The Class-1E
battery system is designed to mitigate the consequences of an
accident, and therefore, can not contribute to the initiation of any
accident. The proposed TS LCO Action Statements will continue to
ensure that the Class-1E batteries are capable of performing their
required safety functions while providing a sufficiently
conservative period of continued plant operation. In addition, this
proposed TS change will not increase the probability of occurrence
of a malfunction of any plant equipment important to safety, since
the manner in which the Class-1E battery system is operated is not
affected by these proposed changes. The operating limits specified
in the proposed TS LCO ensure that the battery's safety functions
will be accomplished. Therefore, the proposed TS changes would not
result in the increase of the consequences
[[Page 12295]]
of an accident previously evaluated, nor do they involve an increase
in the probability 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 TS changes do not involve any physical changes to
the design of any plant SSC. The design and operation of the Class-
1E battery system is not changed from that currently described in
the UFSAR [Updated Final Safety Analysis Report], only the
allocation of battery design margin would be temporarily affected by
the proposed TS LCO. The Class-1E battery system will continue to
function as designed to mitigate the consequences of an accident.
Establishing a 31 day period where a Class-1E battery would be
considered operable, with electrolyte temperature at or above
65 deg.F and Category A and Category B limits met as appropriate,
does not permit plant operation in a configuration that would create
a different type of malfunction to the Class-1E batteries than any
previously evaluated. In addition, the proposed TS changes do not
alter the conclusions described in the UFSAR regarding the safety
related functions of the Class-1E batteries or their support
systems.
Therefore, the proposed TS 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 changes contained in this submittal would implement
TS requirements that either: (1) Permit continued plant operation
when the safety function of the Class-1E batteries can be performed;
or (2) conservatively require placing the plant in a safe shutdown
condition. A Class-1E battery operating within Category A and
Category B limits as appropriate and a 65 deg.F battery electrolyte
temperature (for a limited 31 day period) will still perform its
safety-related functions. Temporary allocation of battery capacity
margins in compensation of degraded operating conditions (low
specific gravity) is already permitted by the Hope Creek TS (for a
31 day period). The ability of the Class-1E batteries to
independently supply their required loads for four hours without
support from battery chargers is not adversely affected by these
proposed changes. Therefore, 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.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant (PBNP), Units 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of amendment request: January 19, 2000 (TSCR 217).
Description of amendment request: The proposed amendments would
revise Technical Specification 15.4.4 to clarify that a different
containment tendon may be designated as a control tendon providing that
the new control tendon had not previously been physically changed.
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 Point Beach Nuclear Plant in accordance with
the proposed amendment does not result in a significant increase in
the probability or consequences of any accident previously
evaluated.
The proposed change does not involve a change to structures,
systems, or components which would affect the probability or
consequences of an accident previously evaluated in the PBNP Final
Safety Analyses Report (FSAR). The containment tendons are
components integral to maintaining the containment pressure boundary
under post accident conditions. Neither the tendons nor the
containment tendon testing process are accident initiators. The
proposed change simply clarifies the Technical Specifications
regarding the selection of control tendons used to develop a tendon
relaxation history and correlate observed test data. The proposed
change does not affect reactor operations or accident analysis and
has no significant radiological consequences. Therefore, this change
will not create a significant increase in the probability or
consequences of an accident previously evaluated.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a new or different kind
of accident from any accident previously evaluated.
The proposed change does not involve a change to the plant
design or operation. As a result, the proposed change does not
affect any of the parameters or conditions that contribute to
initiation of any accidents. This change clarifies the Technical
Specifications regarding the selection of control tendons used to
develop a history and correlate observed test data. Except for the
method of selecting the control tendon, the methods for performing
the actual tendon surveillances are not changed. No new accident
modes are created by selecting the control tendons. No safety-
related equipment or safety functions are altered as a result of
this change. Selecting a control tendon has no influence on, nor
does it contribute to, the possibility of a new or different kind of
accident or malfunction from those previously analyzed. Therefore,
the proposed change will not create the possibility of a new or
different kind of accident previously evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant reduction
in a margin of safety.
The proposed change affects only the selection of control
tendons used to develop a history and correlate observed test data.
Except for the method of selecting the control tendons, the methods
for performing the actual tests are not changed. The proposed change
is based on NRC accepted provisions contained in Regulatory Guide
1.35, Revision 3. Furthermore, the proposed change will not reduce
the availability of systems associated with containment integrity
when they are required to mitigate accident conditions. Therefore,
the proposed change will not create a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Claudia M. Craig.
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
[[Page 12296]]
assessment need be prepared for these amendments. If the Commission has
prepared an environmental assessment under the special circumstances
provision in 10 CFR 51.12(b) and has made a determination based on that
assessment, it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and electronically from the ADAMS Public
Library component on the NRC Web site, http://www.nrc.gov (the
Electronic Reading Room).
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina.
Date of application for amendment: July 9, 1999, as supplemented on
January 19, 2000.
Brief description of amendment: This amendment revises Technical
Specification (TS) \3/4\.2.2, ``Heat Flux Hot Channel Factor--
FQ(Z),'' TS \3/4\.2.3, ``RCS Flow Rate And Nuclear Enthalpy
Rise Hot Channel Factor,'' TS \3/4\.2.5, ``DNB Parameters,'' an
associated note in TS Table 2.2-1, and associated Bases. Specifically,
the proposed amendment would: (1) Remove the allowance for reduced
power operation for reduced Reactor Coolant System (RCS) flow rate
conditions; (2) separate the requirements for F delta H and RCS flow
rate in the format prescribed by NUREG-1431, Revision 1, ``Standard
Technical Specifications, Westinghouse Plants,'' dated April 1995; and
(3) implement the guidance of NUREG-1431, Revision 1, and NRC Generic
Letter 88-16, dated October 4, 1988, for TS \3/4\.2.2 and TS \3/4\.2.3
and associated Bases by removing cycle-specific parameters and placing
that information into the Core Operating Limits Report.
Date of issuance: February 24, 2000.
Effective date: February 24, 2000.
Amendment No. 95.
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 11, 1999 (64 FR
43765).
The January 19, 2000, submittal contained clarifying information
only, and did not change the initial no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated February 24, 2000.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: July 9, 1999.
Brief description of amendment: This amendment revises the
Technical Specifications (TS) by relocating several instrumentation TS
to plant procedures.
Date of issuance: February 24, 2000.
Effective date: February 24, 2000.
Amendment No. 96.
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: August 11, 1999 (64 FR
43766).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 24, 2000.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: July 30, 1999, as supplemented by letter
dated December 2, 1999.
Brief description of amendment: By application dated July 30, 1999,
as supplemented by letter dated December 2, 1999, Entergy Operations,
Inc. (the licensee) requested changes to the Technical Specifications
(TSs) (Appendix A to Facility Operating License No. NPF-47) for the
River Bend Station, Unit 1. The proposed change, more commonly referred
to as ``power uprate,'' would revise the TSs and the operating license
to increase the current licensed power of 2894 megawatts thermal
(MWth) to the uprated power level of 3039 MWth,
an increase of 5 percent. Included in the power uprate license
amendment application was a request to increase the main steam safety
and relief valves (S/RV) safety mode/function setpoint tolerance
defined in Surveillance Requirement (SR) 3.4.4.1 from +0/-2 percent to
3 percent.
This amendment approves, prior-to the issuance of the power uprate
license amendment, a portion of the S/RV setpoint tolerance change
requested. The change increases the safety function lift setpoint
tolerances for the S/RVs listed in SR 3.4.4.1 from the current +0/-2
percent of the safety function lift setpoint to +0/-3 percent (i.e., a
partial 3 percent tolerance). The remaining (``+3 percent'') portion of
the proposed setpoint tolerance change will be reviewed in conjunction
with approval for the power uprate.
Date of issuance: February 9, 2000.
Effective date: As of the date of issuance and shall be implemented
30 days from the date of issuance.
Amendment No.: 109.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 17, 1999.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 9, 2000.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: December 23, 1998.
Brief description of amendment: Modification of Limiting Condition
for Operation for the chlorine detection system and correction of
typographical error in Table 3.3-4.
Date of issuance: February 11, 2000.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 156.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 24, 1999 (64
FR 9190).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 11, 2000.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: March 3, 1999.
Brief description of amendment: The amendment revises Final Safety
Analysis Report, Section 9.5.4.1. The revision changes this section to
explicitly list the Waterford Steam Electric Station, Unit 3 (Waterford
3) deviations from American National Standards Institute (ANSI)
Standard N195-1976.
Date of issuance: February 15, 2000.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
[[Page 12297]]
Amendment No.: 157.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 17, 1999 (64
FR 62713).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 15, 2000.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of application for amendments: December 24, 1998, as
supplemented January 6, 1999.
Brief description of amendments: These amendments change the Beaver
Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and BVPS-2) Technical
Specifications (TSs) to ensure that Emergency Diesel Generator (EDG)
requirements contained in Technical Specification 3/4.8.1 for both
units are consistent with assumptions contained in design analyses and
requirements of plant procedures. Revisions to TS 3/4.8.1, ``A.C.
Sources,'' contained in these amendments provide more conservative
limiting conditions for operation and surveillance requirements that
affect EDG fuel oil storage volume, EDG load rejection and overspeed
testing, and EDG operating frequency requirements. The applicable bases
for each unit are also refined, as necessary, to strengthen the
explanations regarding EDG fuel oil storage systems and provide the EDG
overspeed in terms of frequency (Hertz) and speed (Revolutions Per
Minute).
Date of issuance: February 11, 2000.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 227 and 105.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 27, 1999, (64
FR 4154). The January 6, 1999, letter requested a 60-day implementation
period. This letter did not change the initial proposed no significant
hazards consideration determination or expand the amendments beyond the
scope of the initial notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 11, 2000.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: July 27, 1999.
Brief description of amendment: This amendment eliminates TS 6.4,
``Training,'' and relocates TS 6.5.2.8, ``Audits,'' and TS 6.10,
``Record Retention,'' to the USAR Chapter 17 Quality Assurance Program.
Additionally, the record keeping requirements of TS 6.14, ``Process
Control Program,'' and TS 6.15, ``Offsite Dose Calculation Manual,''
are also being relocated to the USAR Chapter 17 Quality Assurance
Program. Finally, an editorial change has been made to TS 6.8,
``Procedures and Programs.''
Date of issuance: February 14, 2000.
Effective date: Immediately, to be implemented within 120 days.
Amendment No.: 235.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 8, 1999 (64
FR 48863).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 14, 2000.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: November 8, 1999
Brief description of amendment: This amendment relocates Technical
Specification (TS) 6.5.1, Station Review Board, and TS 6.5.2, Company
Nuclear Review Board, to the Davis-Besse Nuclear Power Station Updated
Safety Analysis Report Chapter 17.2, Quality Assurance During the
Operations Phase. These changes are consistent with the recommendations
in NRC Administrative Letter 95-06, ``Relocation of Technical
Specification Administrative Controls Related to Quality Assurance,''
dated December 12, 1995.
Date of issuance: February 14, 2000.
Effective date: Immediately, to be implemented within 120 days.
Amendment No.: 236.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 15, 1999 (64
FR 70087).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 14, 2000.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: June 1, 1999, as supplemented
September 25, 1999.
Brief description of amendments: These amendments revise the St.
Lucie, Units 1 and 2, Technical Specifications (TS) 3.5.2 to allow up
to 7 days to restore an inoperable Low Pressure Safety Injection train
to operable status.
Date of Issuance: February 15, 2000.
Effective Date: February 15, 2000.
Amendment Nos.: 164 and 106.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the TS.
Date of initial notice in Federal Register: June 30, 1999 (64 FR
35206). The supplemental September 25, 1999, letter provided additional
information that did not expand the scope of the amendment request
beyond the initial notice or change the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 15, 2000.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of application for amendment: November 17, 1999 Brief
description of amendment: The amendment revised the technical
specification (TS) surveillance testing of the safety-related
ventilation system charcoal to meet the actions requested in Generic
Letter 99-02, ``Laboratory testing of Nuclear-Grade Activated
Charcoal,'' dated June 3, 1999.
Date of Issuance: February 17, 2000.
Effective Date: February 17, 2000.
Amendment No.: 107.
Facility Operating License No. NPF-16: Amendment revised the TSs.
Date of initial notice in Federal Register: January 12, 2000 (65 FR
1923).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 17, 2000.
No significant hazards consideration comments received: No.
[[Page 12298]]
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: December 28, 1999.
Brief description of amendment: The amendment revised Technical
Specifications Table 4.4.6.1.3-1, ``Reactor Vessel Material
Surveillance Program--Withdrawal Schedule.'' The revised requirement
permits the withdrawal of surveillance capsule number 1 at 8 effective
full-power years (EFPY) instead of the original 10 EFPY.
Date of issuance: February 15, 2000.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 90.
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: January 14, 2000 (65 FR
2443).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 15, 2000.
No significant hazards consideration comments received: No.
Northern States Power Company, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota
Date of application for amendment: December 16, 1999, as
supplemented January 21, 2000.
Brief description of amendment: The amendment revises the Technical
Specification (TS) Safety Limit Minimum Critical Power Ratio (SLMCPR)
values for two recirculation pump and single-loop operation, deletes
cycle specific footnotes, updates the single-loop operation Average
Planar Heat Generation rate limiting values, corrects a typographical
error, and deletes an obsolete reference to Siemens fuel.
Date of issuance: February 16, 2000.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 109.
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 29, 1999 (64
FR 73094).
The January 21, 2000, submittal provided clarifying information
that 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 February 16, 2000.
No significant hazards consideration comments received: No.
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: March 18, 1998.
Brief description of amendments: The amendments revise Bases 3/
4.6.2.1, ``Containment Spray System,'' of the current Technical
Specifications (TSs) and Bases 3.6.6, ``Containment Spray and Cooling
Systems,'' of the improved TSs, to clarify that containment spray is
not required to be actuated during recirculation, but may be actuated
at the discretion of the Technical Support Center. Additionally, the
Bases are clarified to state that the ability to spray containment
using the residual heat removal (RHR) system is demonstrated by opening
the RHR Spray Ring Cross Connect Valve 9003A or B. The Bases are
clarified to state that flow to the spray headers can be established
with only one operable RHR pump by closing the cold leg discharge valve
8809A or B.
Date of issuance: February 9, 2000.
Effective date: February 9, 2000.
Amendment Nos.: Unit 1--139 ; Unit 2--139.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Bases.
Date of initial notice in Federal Register: August 26, 1998 (63 FR
45527).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 9, 2000.
No significant hazards consideration comments received: No.
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 10, 1998, as
supplemented by letter dated November 24, 1999.
Brief description of amendments: These amendments revise Technical
Specification (TS) 3/4.3.2, Table 3.3-5, ``Engineered Safety Features
Response Times,'' of the current TS to add the response times for
closure of the main feedwater regulating valves (MFRVs) and MFRV bypass
valves, and trip of the main feedwater pumps (MFWPs). The change would
also revise TS 3/4.7.1.7 to add a limiting condition for operation,
actions, and surveillance requirements for the MFWP turbine stop
valves, and revise the TS 3/4.7.1.7 actions and surveillance
requirements for the MFRVs, MFRV bypass valves, and main feedwater
isolation valves to be consistent with the NUREG-1431 requirements.
Also, the amendments revise Section 3.7.3 and its associated bases of
the improved Technical Specifications (ITS).
Date of issuance: February 22, 2000.
Effective date: February 22, 2000, to be implemented within 30 days
from the date of issuance.
Amendment Nos.: Unit 1--140; Unit 2--140.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 7, 1998 (63 FR
53954). The November 24, 1999, 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 amendments is contained in a Safety
Evaluation dated February 22, 2000.
No significant hazards consideration comments received: No.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: November 24, 1999, as
supplemented January 13, 2000.
Brief description of amendment: The amendment revises the Technical
Specifications to remove the requirement for partial stroking of the
main steam isolation valves twice-per-week.
Date of issuance: February 24, 2000.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 260.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 29, 1999 (64
FR 73095) The letter of January 13, 2000, provided supplemental
information that did not affect the initial proposed no significant
hazard consideration determination of the original notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 24. 2000.
No significant hazards consideration comments received: No.
[[Page 12299]]
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: January 11, 1999 (PCN-499), as
supplemented November 29, 1999.
Brief description of amendments: The amendments revise Technical
Specification 3.7.6 to change the minimum inventory of water maintained
in the condensate storage tank (T-120) from 280,000 gallons to 360,000
gallons during plant operation Modes 1, 2, and 3.
Date of issuance: February 22, 2000.
Effective date: February 22, 2000, to be implemented within 30 days
of issuance.
Amendment Nos.: Unit 2--162; Unit 3--153.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 18, 2000 (65 FR
2648). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 22, 2000.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: June 8, 1999 (PCN-495).
Brief description of amendments: The amendments modify the
Technical Specifications to (1) reflect that charging flow is not
required to mitigate the effects of design-basis small-break loss-of-
coolant accidents (SBLOCAs), (2) increase the maximum as-found lift
pressure positive tolerance of main steam safety valves from +1 percent
to +2 percent of the setting, and (3) list the ABB Combustion
Engineering Supplement 2 SBLOCA evaluation model as an acceptable
method for determining linear heat rate.
Date of issuance: February 22, 2000.
Effective date: February 22, 2000, to be implemented within 30 days
of issuance.
Amendment Nos.: Unit 2--163; Unit 3--154
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 30, 1999 (64 FR
35210)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 22, 2000.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: January 2, 1998 (PCN-482), as
supplemented December 13, 1999.
Brief description of amendments: The amendments revise Technical
Specification 3.7.5 to add a note that states: The steam driven AFW
[auxiliary feedwater] pump is OPERABLE when running and controlled
manually to support plant start-ups, plant shut-downs, and AFW pump and
valve testing.
Date of issuance: February 23, 2000.
Effective date: February 23, 2000, to be implemented within 30 days
of issuance.
Amendment Nos.: Unit 2--164; Unit 3--155.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register January 19, 2000 (65 FR
2991), as corrected January 26, 2000 (65 FR 4265)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 23, 2000.
No significant hazards consideration comments received: No
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: March 19, 1999.
Brief description of amendments: These amendments relocate
Technical Specification Section 3/4.8.3, ``Electrical Equipment
Protective Devices,'' and the associated bases to the Technical
Requirements Manual.
Date of issuance: February 22, 2000.
Effective date: As of date of issuance to be implemented no later
than 45 days after issuance.
Amendment Nos.: 250 and 241.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: April 21, 1999 (64 FR
19566).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 22, 2000.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 1st day of March 2000.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 00-5477 Filed 3-7-00; 8:45 am]
BILLING CODE 7590-01-P