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Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations
Note: EPA no longer updates this information, but it may be useful as a reference or resource.
[Federal Register: January 24, 2001 (Volume 66, Number 16)]
[Notices]
[Page 7667-7690]
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
[DOCID:fr24ja01-82]
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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 January 2, 2001, through January 12, 2001.
The last biweekly notice was published on January 10, 2001 (66 FR
2010).
Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC's Public Document
Room, located at One White Flint North, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. The filing of requests for a hearing
and petitions for leave to intervene is discussed below.
By February 23, 2001, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, located at One
White Flint North, 11555 Rockville Pike (first Floor), Rockville,
Maryland 20852. Publicly available records will be accessible and
electronically from the ADAMS Public Library component on the NRC Web
site, http://www.nrc.gov (the Electronic Reading Room). If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the
[[Page 7668]]
petitioner's interest. The petition should also identify the specific
aspect(s) of the subject matter of the proceeding as to which
petitioner wishes to intervene. Any person who has filed a petition for
leave to intervene or who has been admitted as a party may amend the
petition without requesting leave of the Board up to 15 days prior to
the first prehearing conference scheduled in the proceeding, but such
an amended petition must satisfy the specificity requirements described
above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, located at One White Flint North, 11555 Rockville Pike
(first floor), Rockville, Maryland 20852, by the above date. A copy of
the petition should also be sent to the Office of the General Counsel,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to
the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852.
Publicly available records will be accessible electronically from the
ADAMS Public Library component on the NRC Web site, http://www.nrc.gov
(the Electronic Reading Room).
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: December 29, 2000.
Description of amendment request: The proposed amendment would
increase the Technical Specification allowed outage time from 3 days to
14 days for a single inoperable Division 1 or 2 diesel generator.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
The proposed Technical Specification (TS) changes revise the
Completion Time for Required Actions A.2 and B.4 associated with the
Division 1 and Division 2 Diesel Generators (DG). The proposed
changes allow an extension of the current TS Completion Time from 72
hours to 14 days when the Division 1 or Division 2 DG is inoperable.
The proposed changes do not affect the design of the DGs, the
operational characteristics of function of the DGs, the interfaces
between the DGs and other plant systems, or the reliability of the
DGs. Required Actions and the associated Completion Times are not
initiating conditions for any accident previously evaluated, and the
DGs are not initiators of any previously evaluated accidents. The
DGs mitigate the consequences of previously evaluated accidents
including a loss of offsite power. The consequences of a previously
analyzed event will not be significantly affected by the extended DG
Completion Time since the DGs will continue to be capable of
performing their accident mitigation function as assumed in the
accident analysis. Thus the consequences of accidents previously
analyzed are unchanged between the existing TS requirements and the
proposed changes. The consequences of an accident are independent of
the time the DGs are out of service as long as adequate DG
availability is assumed. The proposed changes will not result in a
significant decrease in DG availability so that the assumptions
regarding DG availability are not impacted.
To fully evaluate the effect of the proposed EDG Completion Time
extension, Probabilistic Risk Assessment (PRA) methods and a
deterministic analysis were utilized. The results of the analysis
show no significant increase in Core Damage Frequency (CDF) and
Large Early Release Frequency (LERF). Therefore, the proposed
changes do not involve significant increase in the probability or
consequences of an accident previously analyzed.
2. The proposed change would not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve a change in the design,
configuration, or method of operation of the plant. The proposed
changes will not alter the manner in which equipment operation is
initiated, nor will the function demands on credited equipment be
changed. The changes do not alter assumptions made in the safety
analysis. No alteration in the procedures, which ensure that the
plant remains within analyzed limits, is being proposed, and no
changes are being made to the procedures relied upon to respond to
an off-normal event. As such, no new failure modes are being
introduced. Therefore, these proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change will not involve a significant reduction
in the margin of safety.
Since there are no changes to the plant design and safety
analysis, and no changes to the DG design, including any instrument
setpoints, no margin of safety assumed in the
[[Page 7669]]
safety analysis is affected. If a margin of safety is ascribed to DG
availability and plant risk, it has also been determined that such a
margin of safety is not significantly reduced, as the proposed
changes have been evaluated both deterministically and using a risk-
informed approach. The evaluation concluded the following with
respect to the proposed changes.
Applicable regulatory requirements will continue to be met,
adequate defense-in-depth will be maintained, sufficient safety
margins will be maintained, and any increases in CDF and LERF are
small and consistent with the NRC Safety Goal Policy Statement
(Federal Register, Vol. 51, p. 30028 (51 FR 30028), August 4, 1986,
as interpreted by NRC Regulatory Guides 1.174 and 1.177).
Furthermore, increases in risk posed by potential combinations of
equipment out of service during the proposed DG extended Completions
Time will be managed under a configuration risk management program
consistent with 10 CFR 50.65, ``Requirements for Monitoring the
Effectiveness of Maintenance at Nuclear Power Plants,'' paragraph
(a)(4). The following are examples.
An extended DG Completion Time will not be entered
intentionally for scheduled maintenance purposes if severe weather
conditions are expected.
While in the extended DG Completion Time, additional
elective equipment maintenance or testing or equipment failure will
be evaluated. Activities that yield unacceptable results will be
avoided.
The condition of the offsite power supply and
switchyard will be evaluated.
Activities have been identified that can mitigate any
increase in risk. Procedures are in place for the minimizing risk
associated with the following activities:
No elective maintenance will be scheduled within the switchyard
that would challenge the offsite power connection or offsite power
availability during the extended DG Completion Time.
No elective work will be performed on protected equipment or
opposite train emergency core cooling system (ECCS) equipment during
the extended DG Completion Time.
The availability of offsite power coupled with the availability
of the other DGs and the use of on-lime risk assessment tools
provide adequate compensation for the potential small incremental
increase in plant risk of the extended DG Completion Time. In
addition, the increased availability of the DGs during refueling
outages offsets the small increase in plant risk during operation.
The proposed extended DG Completion Times in conjunction with the
availability of the other DGs continues to provide adequate
assurance of the capability to provide power to the engineered
safety features (ESF) buses. Therefore, implementation of the
proposed changes will 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: Kevin P. Gallen, Morgan, Lewis & Bockius,
LLP, 1800 M Street, NW, Washington, DC 20036-5869.
NRC Section Chief: Anthony J. Mendiola.
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of amendment request: December 6, 2000.
Description of amendment request: The proposed amendment requests
changes to the once-through steam generator tube inspection criteria in
order to allow certain inside diameter inter-granular attack
indications to remain in service. This amendment request seeks to make
permanent the tube inspection criteria that have been used for the past
two operating cycles at TMI-1.
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. Operation of the facility in accordance with the proposed
amendment will not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed flaw disposition strategy, based on measurable eddy
current parameters of axial and circumferential extent for Inside
Diameter (ID) Initiated Inter-Granular Attack (IGA), will continue
to provide high confidence that unacceptable flaws that do not have
the required structural integrity to withstand a postulated MSLB
[main steam line break] are removed from service. The axial and
circumferential length limits for eddy current ID degradation
indications meet Draft Regulatory Guide 1.121 (Reference 9 [of the
licensee's application]) acceptance criteria for margin to failure
for MSLB-applied differential pressure and axial tube loads. The
capability for detection of flaws is unaffected; and the
identification of tubes that should be repaired or removed from
service is maintained. The operation of the OTSGs [once-through
steam generators] or related structures, systems, or components is
otherwise unaffected. Therefore, neither the probability nor
consequences of a Steam Generator Tube Rupture (SGTR) is
significantly increased either during normal operation or due to
limiting loads of a MSLB accident.
Therefore, operation of the facility in accordance with the
changes included in LCA [license change application] No. 291 will
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
B. Operation of the facility in accordance with the proposed
amendment will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
because there are no hardware changes involved nor changes to any
operating practices. These changes involve only the OTSG tube
inservice inspection surveillance requirements, which could only
affect the potential for OTSG primary-to-secondary leakage which has
been analyzed and is subject to Technical Specification requirements
not affected by these changes. The proposed changes continue to
impose flaw length limits for ID IGA to assure tube structural and
leakage integrity.
Therefore, operation of the facility in accordance with the
changes included with LCA No. 291 will not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
C. Operation of the facility in accordance with the proposed
amendment will not involve a significant reduction in a margin of
safety.
The margins of safety defined in Draft Regulatory Guide 1.121
(Reference 9 [of the licensee's application]) are retained. The
probability of detecting degradation is unchanged since the bobbin
coil eddy current methods will continue to be the primary means of
initial detection and the probability of leakage from any
indications left in service remains acceptably small. The strategy
of dispositioning ID-initiated IGA indications will continue to
provide a high level of confidence that tubes exceeding the
allowable limits for tube integrity are repaired or removed from
service.
Therefore, operation in accordance with the changes included in
LCA No. 291 will not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Edward J. Cullen, Jr., Esq., PECO Energy
Company, 2301 Market Street, S23-1, Philadelphia, PA 19103.
NRC Section Chief: Marsha Gamberoni.
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of amendment request: December 6, 2000.
Description of amendment request: The proposed amendment provides
clarifications to the decay heat removal (DHR) Technical Specifications
(TSs). It is intended, in part, to fulfill a
[[Page 7670]]
commitment made by the licensee to the NRC during a pre-decisional
enforcement conference on April 23, 1999. Specifically, the proposed
changes would: (1) Define and clarify the emergency feedwater (EFW)
flowpath redundancy as described in the Bases; (2) provide operability
requirements for the redundant steam supply paths to the turbine-driven
EFW pump; (3) provide a more conservative 72-hour allowed outage time
(AOT) with any EFW pump or flowpath inoperable; (4) provide a more
conservative 1-hour AOT with both EFW flowpaths to a single once-
through steam generator (OTSG) inoperable or with 2 EFW pumps
inoperable; and (5) revise and clarify EFW pump and flowpath
operability requirements during surveillance testing. Minor
administrative and editorial changes are also proposed. A change to the
Bases for TS 3.5.5, ``Accident Monitoring Instrumentation,'' regarding
the description of the pressurizer level instrument channels to reflect
the replacement of Bailey transmitters was also included.
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. Operation of the facility in accordance with the proposed
amendment will not involve a significant increase in the probability
or consequences of an accident previously evaluated.
This change incorporates the concept of EFW flowpath redundancy
thoughout the TS[s], which takes into consideration the redundancy
provided by the EFW System modifications made in the mid-1980s after
the accident at TMI-2. This change incorporates a 72 hour required
action time when redundant components are made inoperable. These
changes do not result in any change to the configuration of the EFW
System as described in the [UF]SAR [Updated Final Safety Analysis
Report] or used in plant specific analyses. The reliability of EFW
System components is unaffected. The 72 hour required action time
for inoperability of redundant EFW components ensures that the EFW
System can fulfill its safety function to provide adequate OTSG
cooling during a design basis accident (DBA). The one hour required
action time ensures prompt action to initiate a plant shutdown when
the design flow capability of the EFW System cannot be assured.
The current TS 4.9.1.2 contains EFW flowpath operability
requirements during surveillance testing rather than requiring that
a specific test be performed as do the other subparagraphs of TS
4.9.1. For this reason the requirements of TS 4.9.1.2 are being
moved to the LCO [limiting condition for operation] section in
Chapter 3 and combined with the note following the current TS
3.4.1.1.a(2) into a new TS 3.4.1.1.a(4) to define the EFW System
operability requirements for EFW pumps and flowpaths during
surveillance testing. The new specification incorporates the
consideration of EFW flowpath redundancy consistent with HSPS [Heat
Sink Protection System] train operability requirements and continues
to require that compensatory measures be implemented to promptly
restore components if EFW is needed during surveillance testing when
more than one flowpath is made inoperable to an OTSG. The intent of
this surveillance standard has been retained, which assures that the
minimum number of EFW flowpaths to the OTSGs will be available with
minimal operator action.
This change provides further assurance that EFW System design
basis requirements will be met and does not affect EFW System
configuration, setpoints, or reliability. These changes will not
affect any accident initiation sequence and do not affect off site
dose consequences of accidents that have been analyzed.
The editorial changes included in this LCA [license change
application] are intended to improve the clarity, consistency, and
reliability of the TS[s] [and] do not change the intent or
interpretation.
Therefore, operation of the facility in accordance with the
changes included in LCA-286 will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
B. Operation of the facility in accordance with the proposed
amendment will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
As a result of this change, no additional hardware is being
added; and there will be no effect on EFW System design, operation
as described in the [UF]SAR, or assumptions used in plant specific
analyses. The requirement for three EFW Pumps and [associated]
flowpaths to be operable for continuous plant operation is not
affected by this change. Events involving the EFW System operation
have been reviewed and determined to have no impact from these
changes. The additional operability requirements for the turbine-
driven EFW Pump steam supplies, the revised LCOs [limiting condition
for operation], and changes to define EFW flowpath redundancy
ensures minimum EFW component operability as credited in plant
analyses. The editorial changes included in this LCA are intended to
improve clarity, consistency and readability of the TS[s] and Bases,
[and] do not change the intent or interpretation.
Therefore, operation of the facility in accordance with the
changes included with LCA-286 will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
C. Operation of the facility in accordance with the proposed
amendment will not involve a significant reduction in a margin of
safety.
This change does not affect the EFW System design or
instrumentation setpoints. The requirement for three operable EFW
pumps and associated flowpaths is not affected by this change. The
revised LCO imposes a 72 hour required action time when any EFW pump
or redundant flowpath to either OTSG is inoperable, including
inoperability for the purpose of conducting surveillance testing.
The revised LCO requires that at least one flowpath to each OTSG
must be operable or a plant shutdown is required to be initiated
within one hour. The 8 hour action time currently allowed for pump
inoperability during surveillance testing is also applied to
flowpath inoperability during testing. The revised LCO continues to
require compensatory measures during EFW testing when HSPS [heat
sink protection system] is required to be operable and an OTSG is
isolated, retaining the provision that EFW flowpath valves can be
realligned promptly from their test mode to their operational
allignment if EFW flow is needed. The revised Accident Monitoring
Instrumentation specification is needed to reflect the revised
flowpath definition and does not change the intent of the
specification. The editorial changes included in this LCA are
intended to improve the clarity, consistency, and readability of the
TS[s] [and] do not change the intent or interpretation.
Therefore, operation in accordance with the changes included in
LCA-286 will not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Edward J. Cullen, Jr., Esq., PECO Energy
Company, 2301 Market Street, S23-1, Philadelphia, PA 19103.
NRC Section Chief: Marsha Gamberoni.
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, 2000.
Description of amendments request: The proposed amendments would
revise the value of the minimum departure from nucleate boiling ratio
(DNBR) from `` 1.30'' in the current technical
specifications to `` 1.3 (through operating cycle 10)'' and
`` 1.34 (operating cycle 11 and later)'' in the safety
limits Technical Specification (TS) 2.1.1.1 and in function 15, DNBR--
Low, in Table 3.3.1-1, ``Reactor Protective System Instrumentation.''
The proposed amendments are structured such that the ``
1.34'' would become effective for each unit in operating cycle 11 and
later. Operating cycle 11 begins in spring 2002 for Unit
[[Page 7671]]
2, in fall 2002 for Unit 1, and in spring 2003 for Unit 3. From now to
operating cycle 11, the `` 1.30'' will remain the minimum
DNBR requirement for the three units.
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 purpose of the proposed Technical Specification (TS)
amendment is to provide a revised Departure from Nucleate Boiling
Ratio (DNBR) Safety Limit (TS Section 2.1.1.1) and Low DNBR Reactor
Protective System (RPS) trip setpoint (TS Limiting Condition for
Operation (LCO) 3.3.1, Table 3.3.1-1).
The proposed TS amendment involves increasing the DNBR Safety
Limit and Low DNBR RPS trip setpoint from `` 1.30'' to
`` 1.34''. Changing this limit in and of itself will not
alter the physical characteristics of any component involved in the
initiation of an accident. Thus, the proposed change does not
involve a significant increase in the probability of an accident
previously evaluated.
The Core Operating Limit Supervisory System (COLSS) Power
Operating Limit (POL) is an alarm limit on the maximum steady state
core power level. The alarm is based on maintaining COLSS calculated
DNBR a pre-determined amount above the DNBR Safety Limit. The Low
DNBR RPS trip setpoint[,] in conjunction with the COLSS POL,
prevents the DNBR in the limiting coolant channel in the core from
violating the DNBR Safety Limit during design basis Anticipated
Operational Occurrences (AOO). Operating below the COLSS POL ensures
the Low DNBR RPS trip setpoint will protect the core [fuel] from
damage due to the occurrence of locally saturated conditions in the
limiting (hot) channel during the worst AOO. Thus, during normal and
anticipated operation the Low DNBR RPS trip setpoint in conjunction
with the COLSS POL prevents overheating of the fuel cladding and
subsequent cladding perforation that would release fission products
to the reactor coolant.
This change will accommodate increased DNBR sensitivity to
uncertainties in inlet flow to the hot assembly and adjacent
assemblies. This increased sensitivity is attributed to the flatter
power distributions of the more efficient present day erbium core
designs. More adverse DNBR sensitivity to inlet flow was first
encountered in Unit 1 Cycle 7. At that time the increased DNBR
sensitivity was accounted for statistically by applying a thermal
margin penalty to Core Operating Limit Supervisory System (COLSS)
and Core Protection Calculators (CPCs) using approved Statistical
Combination of Uncertainties (SCU) methods. This approach was also
used for the subsequent cycles in all units up until the present.
The NRC Safety Evaluation (issued May 26, 1994 for Palo Verde
Nuclear Generating Stations (PVNGS) Units 1, 2, and 3) for the
present `` 1.30'' DNBR limit states, ``Uncertainties in
inlet flow to the hot assembly and adjacent assemblies can be
accounted for statistically by either increasing DNBR or applying a
thermal margin penalty using approved SCU methods.''
The proposed TS amendment change for DNBR Safety Limit and Low
DNBR RPS trip setpoint limit ( 1.34) was calculated using
approved SCU methods to statistically include the above described
increased DNBR sensitivity. This new DNBR limit was calculated such
that it has a high probability of covering all future cycle designs.
Thus, this change involves moving the existing increased inlet flow
uncertainty penalty from a thermal margin penalty contained within
COLSS and CPCs to an increase in the DNBR Safety Limit and Low DNBR
RPS trip setpoint limit. The DNBR Safety Limit and Low DNBR RPS trip
setpoint increases from `` 1.30'' to `` 1.34''
due to this change. The COLSS and CPCs would respond similarly with
the increased inlet flow uncertainty penalty located in either the
COLSS or CPCs or in the DNBR Safety Limit. The proposed amendment
changes only the location of the increased inlet flow uncertainty
penalty and does not impact the operation of the plant. The core
power distribution during all phases of normal and anticipated
operational occurrences will remain bounded by the initial
conditions assumed in Chapter 15 of the Palo Verde Nuclear
Generating Station (PVNGS) UFSAR [Updated Final Safety Analysis
Report]. Thus, the proposed change does not involve a significant
increase in 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.
This change does not alter the physical design of any System,
Structure, or Component (SSC) of the plant.
The change involves increasing the DNBR Safety Limit and the Low
DNBR RPS trip setpoint from `` 1.30'' to ``
1.34'' and decreasing the corresponding DNBR thermal margin penalty
factors in COLSS and CPC in a compensating manner. Changing these
limits and penalty factors will not alter the physical or functional
characteristics of any component in the plant. These changes will
not affect any safety-related equipment used in the mitigation of
anticipated operational occurrences or design basis accidents. Thus,
the proposed change 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 DNBR Safety Limit specified in Section 2.1.1.1 and the Low
DNBR RPS trip setpoint specified in Table 3.3.1-1 of LCO 3.3.1 of
[the] PVNGS Technical Specifications ensure that operation of the
reactor does not result in a departure from nucleate boiling during
normal operation and design basis anticipated operational
occurrences. Therefore, operating consistent with the increased DNBR
Safety Limit and Low DNBR RPS trip setpoint will ensure that no
anticipated operational occurrences will result in core conditions
below the specified DNBR Safety Limit and no postulated accident
exceeds the site boundary dose limits. The UFSAR Chapter 15 analysis
remains bounding and the margins of safety will be maintained
because the COLSS and the CPC overall uncertainty factors will be
calculated and implemented consistent with the increased DNBR Safety
Limit of `` 1.34''. Therefore, this change to TS Section
2.1.1.1 and Table 3.3.1-1 of LCO 3.3.1 does not involve a
significant reduction in a margin of safety.
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 5, 2000.
Description of amendments request: The proposed amendments would
revise the action statement for Specification 3.7.5, ``Auxiliary
Feedwater (AFW) System,'' of the Technical Specifications (TSs). The
amendments would incorporate NRC-approved TS Task Force (TSTF) Traveler
Number TSTF-340, Revision 3, to allow a 7-day Completion Time for the
turbine-driven AFW pump if inoperability occurs in reactor Mode 3
following a refueling outage, and if Mode 2 had not been entered.
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 proposed change involve a significant increase in
the probability or
[[Page 7672]]
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 amendment to Technical Specification 3.7.5 would
allow a 7 day Completion Time for Condition A for the turbine-driven
Auxiliary Feedwater (AFW) pump if the inoperability occurs in MODE 3
following a refueling outage, if MODE 2 had not been entered.
Extending the Completion Time does not involve a significant
increase in the probability or consequences of an accident
previously evaluated because: (1) The proposed amendment does not
represent a change to the system design, (2) the proposed amendment
does not prevent the safety function of the AFW system from being
performed since the other fully redundant esstential train and the
non-essential train are required to be operable, (3) the proposed
amendment does not alter, degrade, or prevent action described or
assumed in any accident described in the PVNGS [Palo Verde Nuclear
Generating Station] UFSAR [Updated Final Safety Analysis Report]
from being performed since the other trains of AFW are required to
be operable, (4) the proposed amendment does not alter any
assumptions previously made in evaluating radiological consequences,
and (5) the proposed amendment does not affect the integrity of any
fission product barrier. No other safety related equipment is
affected by the proposed change. Therefore, this proposed amendment
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed amendment to Technical Specification 3.7.5 would
allow a 7 day Completion Time for Condition A for the turbine-driven
Auxiliary Feedwater (AFW) pump if the inoperability occurs in MODE 3
following a refueling outage, if MODE 2 had not been entered.
Extending the Completion Time does not create the possibility of a
new or different kind of accident from any accident previously
evaluated because: (1) The proposed amendment does not represent a
change to the system design, (2) the proposed amendment does not
alter how equipment is operated or the ability of the system to
deliver the required AFW flow, and (3) the proposed amendment does
not affect any other safety related equipment. 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 proposed change involve a significant reduction in a
margin of safety?
No. The proposed amendment does not involve a significant
reduction in a margin of safety.
The PVNGS safety analysis credits essential Auxiliary Feedwater
(AFW) pump delivery of 650 gpm at a steam generator pressure of 1270
psia or equivalent at the steam generator entrance for design basis
accidents. The AFW System Design Basis Manual (AF), Revision 11,
states that these pumps are designed to supply 750 gpm. The proposed
[***] amendment to Technical Specification 3.7.5 would allow a 7 day
Completion Time for Condition A for the turbine-driven AFW pump if
the inoperability occurs in MODE 3 following a refueling outage, if
MODE 2 had not been entered. Extending the Completion Time does not
involve a significant reduction in a margin of safety because: (1)
During a return to power operations following a refueling outage,
decay heat [in the core] is at its lowest levels, (2) the other
essential and non-essential AFW trains are required to be OPERABLE
when MODE 3 is entered, (3) the essential motor-driven AFW train can
provide sufficient flow to remove decay heat and cool the unit to
Shutdown Cooling system entry conditions from power operations, and
4) the non-essential motor-driven AFW train is designed to supply
sufficient water to remove decay heat with steam generator pressure
at no load conditions to cool the unit to Shutdown Cooling entry
conditions.
Based on the responses to these three criteria, APS [Arizona
Public Service Company] has concluded that the proposed amendment
involves no significant hazards considerations.
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.
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 amendment request: December 13, 2000
Description of amendment request: The proposed amendment would
revise Harris Nuclear Plant (HNP) Technical Specification (TS) 3/4.9.2
``Refueling Operations--Instrumentation'' and the associated Bases to
permit using alternate installed detectors or temporary source range
detectors instead of the two Source Range Nuclear Flux Monitors
specified in the current HNP TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
This change only involves reactor core monitoring requirements
during Mode 6. These monitoring requirements are not credited for
accident mitigation. Alternate monitors will be provided with the
accuracy and sensitivity required to adequately monitor changes in
the core reactivity levels during refueling activities. Neutron Flux
monitors are for indication only and do not interface with other
structures, systems, or components that might initiate an accident.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Neutron Flux monitors are for indication only and do not
interface with other structures, systems, or components that might
initiate an accident. The proposed change will not modify plant
systems or operate plant components such that a new or different
accident scenario is created.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
Similar changes, to the proposed change, have been approved at
the Beaver Valley Power Station and the Diablo Canyon Power Plant.
The proposed change will maintain adequate monitoring of core
reactivity in Mode 6. The proposed change maintains requirements for
two operable neutron flux monitors. Neutron flux monitors are not
credited in the HNP accident analyses for accident mitigation in
Mode 6.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Richard P. Correia.
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 amendment request: December 14, 2000.
[[Page 7673]]
Description of amendment request: The proposed amendment would
revise Harris Nuclear Plant (HNP) Technical Specification (TS) 3/4.8.1
related to emergency diesel generators (EDGs). Specifically, the
licensee proposes revising TS Surveillance Requirement 4.8.1.1.2.f.7,
the 24-hour EDG endurance run test, by removing the restriction to
perform the test during shutdown conditions.
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 amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The EDGs and their associated emergency buses are not accident
initiating equipment; therefore, there will be no impact on accident
probabilities due to this proposed amendment. The EDGs mitigate the
consequences of previously evaluated accidents involving a loss of
offsite power. The proposed amendment continues to assure the EDGs
perform their function when called upon. The design of the equipment
is not being modified. The proposed amendment does not impact the
operational characteristics of the EDGs, the interfaces between the
EDGs and other plant systems, or the function or reliability of the
EDGs. The EDGs remain capable of performing their accident
mitigation function. The HNP Probabilistic Safety Analysis (PSA)
model results are not affected by the proposed change.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed amendment does not alter the design, configuration,
or method of operation of the plant. No physical changes are being
proposed, nor any changes to the method of operation of the EDGs or
supporting systems. The proposed amendment, in effect, allows a
small increase in the duration that the EDGs are operated parallel
to the grid for test purposes. No new system interactions are
created, and the proposed change does not introduce a new failure
mode.
Therefore the proposed change does not create the possibility of
a new or different kind of accident.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed change does not affect the Limiting Conditions for
Operation or their Bases that are used to establish any margin of
safety. The ability of the EDGs to separate from the offsite power
source has been designed and tested per Technical Specification
requirements. The proposed change does not involve a change to the
plant design or operation and does not affect the availability of
any of the required power sources, nor the capability of the EDGs to
perform their intended safety function.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The foregoing analysis demonstrates that the proposed amendment
to HNP TS does not involve a significant increase in the probability
or consequences of a previously evaluated accident, does not create
the possibility of a new or different kind of accident, and does not
involve a significant reduction in a margin of safety.
Based upon the preceding analysis, [Carolina Power & Light
Company] CP&L concludes that the proposed amendment does not involve
a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Richard P. Correia.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: December 11, 2000.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) 3.1.F.2.a, ``Primary to Secondary
Leakage,'' and 4.13.A.3.f, ``Steam Generator Tube Inservice
Surveillance,'' based on the prior replacement of the steam generators
(SGs). Specifically, the proposed changes would (1) revise the primary
to secondary leakage limits and (2) delete requirements associated with
tube sleeve repair, steam generator tube denting, F* repair
classification and criteria, and (3) modify the associated TS Bases. In
addition, the proposed amendment includes several related
administrative changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Changes to SG Primary to Secondary Leakage Limits
1. Does the change involve a significant increase in the
probability [* * *] or consequences of an accident previously
evaluated?
The proposed reduction in primary to secondary leakage limit and
the elimination of the limit for SGs containing sleeved tubes does
not affect accident initiators or precursors. The proposed change
establishes a primary to secondary leakage limit that is equivalent
to the lesser of the primary to secondary leakage limits currently
established for SG with and without SG tube sleeves. Reducing the
primary to secondary leakage limit does not increase the probability
of an accident. The proposed change does not increase primary to
secondary leakage limits. Therefore, the consequences of an accident
are not increased. Therefore, the probability of occurrence or the
consequences of accidents previously evaluated are not significantly
increased.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not modify any plant equipment.
Therefore, the proposed changes do not degrade the reliability of
systems, structures, or components or create a new accident
initiator or precursor. No new failure modes are created. Therefore,
the change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed change establishes one limit for primary to
secondary limit that is the same as the most restrictive of the two
primary to secondary leakage limits that currently exists. The
proposed change does not increase the allowable primary to secondary
leakage limit.
Since the primary to secondary leakage limit is not increased,
the margin of safety will not be reduced. The proposed change still
requires verification that primary to secondary leakage is within
the limit at the existing frequency. Since the primary to secondary
leakage limit is not increased, dose rates at the site boundary will
not be increased. Therefore, the proposed activity does not involve
a significant reduction in a margin of safety.
Deletion of Provisions Associated With SG Tube Sleeving Repair
Method
1. Does the change involve a significant increase in the
probability [* * *] or consequences of an accident previously
evaluated?
The proposed deletion of the SG tube sleeving provisions does
not affect accident initiators or precursors. The proposed change
deletes the TS provisions that are not approved for the replacement
SGs. Deletion of an unapproved repair method from the TS does not
increase the probability of an accident and the proposed change does
not increase primary to secondary leakage limits. Consequently, the
consequences of an accident are not significantly increased.
Therefore, the probability of occurrence or
[[Page 7674]]
the consequences of accidents previously evaluated are not
significantly increased.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not impact or interface with plant
safety related equipment. Therefore, the proposed changes do not
degrade the reliability of systems, structures, or components or
create a new accident initiator or precursor. No new failure modes
are created. Therefore, the change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed change deletes the TS provisions that are not
approved for the replacement SGs. The proposed change does not
increase the allowable primary to secondary leakage limit. Since the
primary to secondary leakage limit is not increased, the margin of
safety will not be reduced. Therefore, the proposed activity does
not involve a significant reduction in a margin of safety.
Deletion of Provisions Associated with Steam Generator F* Tube
Classification
1. Does the change involve a significant increase in the
probability [* * *] or consequences of an accident previously
evaluated?
The proposed deletion of the F* criteria and associated
provisions does not affect accident initiators or precursors. The
proposed change deletes the TS provisions that are not approved for
the replacement SGs. Deletion of an unapproved repair method from
the TS does not increase the probability of an accident. The
proposed change does not increase primary to secondary leakage
limits. Therefore, the probability of occurrence or the consequences
of accidents previously evaluated are not significantly increased.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not impact or interface with plant
safety related equipment. Therefore, the proposed changes do not
degrade the reliability of systems, structures, or components or
create a new accident initiator or precursor. No new failure modes
are created. Therefore, the change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed change deletes the TS provisions that are not
approved for the replacement SGs. The proposed change does not
increase the allowable primary to secondary leakage limit. Since the
primary to secondary leakage limit is not increased, the margin of
safety will not be reduced. Therefore, the proposed activity does
not involve a significant reduction in a margin of safety.
Deletion of Provisions Associated With SG Tube Denting Phenomenon
1. Does the change involve a significant increase in the
probability [* * *] or consequences of an accident previously
evaluated?
The proposed deletion of the requirements and associated
provisions regarding SG tube denting does not significantly affect
accident initiators or precursors. The proposed change deletes from
the TS provisions that are not necessary for the replacement SGs.
Deletion of the SG tube denting examination requirements from the TS
does not increase the probability of an accident. The proposed
change does not increase primary to secondary limits. Therefore, the
consequences of an accident are not increased. Therefore, the
probability of occurrence or the consequences of accidents
previously evaluated are not significantly increased.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not impact or interface with plant
safety related equipment. Therefore, the proposed changes do not
degrade the reliability of systems, structures, or components or
create a new accident initiator or precursor. No new failure modes
are created. Therefore, the change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed change deletes the TS provisions that are not
applicable for the replacement SGs. The proposed change does not
increase the allowable primary to secondary leakage limit. Since the
primary to secondary leakage limit is not increased, the margin of
safety will not be reduced. Therefore, the proposed activity does
not involve a significant reduction in a margin of safety.
Related Administrative Changes
1. Does the change involve a significant increase in the
probability [* * *] or consequences of an accident previously
evaluated?
The proposed administrative changes do not affect accident
initiators or precursors. The proposed changes correct the
presentation of several TS Basis pages and delete an obsolete
scheduler extension footnote. Correcting the page presentation and
deleting an obsolete footnote do not increase the probability of an
accident. The proposed change does not increase primary to secondary
leakage limits. Consequently, the consequences of an accident are
not significantly increased.
Therefore, the probability of occurrence or the consequences of
accidents previously evaluated are not significantly increased.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not impact or interface with plant
safety related equipment. Therefore, the proposed changes do not
degrade the reliability of systems, structures, or components or
create a new accident initiator or precursor. No new failure modes
are created. Therefore, the change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed administrative changes do not affect accident
initiators or precursors. The proposed change corrects the
presentation of several TS Basis pages and deletes an obsolete
scheduler extension footnote. The proposed changes do not increase
the allowable primary to secondary leakage limit. Since the primary
to secondary leakage limit is not increased, the margin of safety
will not be reduced. Therefore, the proposed activity 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: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Section Chief: Marsha Gamberoni.
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of amendment request: December 7, 2000.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) regarding the Limiting
Conditions for Operation (LCO) for the auxiliary feedwater system (LCO
3.7.5) to be similar to changes to the ``Standard Technical
Specifications, Combustion Engineering Plants,'' NUREG 1432, Revision 1
(STS), made by the Nuclear Energy Institute Technical Specifications
Task Force (TSTF) change number 325, ``Changes To Structure Of
[Emergency Core Cooling System] ECCS--Operating LCO.''
Palisades LCO 3.7.5, ``Auxiliary Feedwater System,'' would be
changed as follows: (1) An editorial change would be made to Note 2 to
put the word ``operable'' in uppercase letters; (2) the second and
third parts of the Condition A description, ``AND--At least 100% of the
required AFW flow available to each steam generator--AND--At least two
AFW pumps OPERABLE,'' would be deleted; (3) the second part of the
Condition B description, ``One or more AFT trains inoperable for
reasons other than Condition A with at least 100% of the required AFW
flow available in MODE 1, 2, or 3,'' would be replaced with two new
parts (``Less than 100% of the required AFW flow available to either
steam generator--OR--Fewer than two AFW pumps OPERABLE in mode 1, 2, OR
3''); and (4) the wording of
[[Page 7675]]
Condition C would be revised to address the condition where
insufficient AFW flow is available to achieve a plant shutdown while in
any mode within the applicable conditions of LCO 3.7.5. The licensee
also forwarded related changes to the TS Bases.
Additional changes requested in the licensee's application dated
December 7, 2000, are based upon other TSTFs and are addressed by
separate Federal Register notices.
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 would not] involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Changes are proposed to LCO 3.7.5, Auxiliary Feedwater, which
emulate changes made to Standard Technical Specifications,
Combustion Engineering Plants, NUREG 1432, Rev.1 (STS) by TSTF 325.
The structure of LCO 3.7.5 has been rearranged to maintain Condition
A (and, in certain circumstances, Condition B) in effect if failures
should occur which reduce available flow to less than 100% of the
required flow (that flow assumed in the accident analyses). The
resulting requirements are those intended when the LCO was initially
constructed and represent the way the LCO Conditions are being
applied. Therefore there is no change in intent or application of
the LCO. In the case where inoperable AFW train components reduce
available flow below that required, and a subsequent partial
restoration is made to provide 100% of the required flow, the
proposed change makes the literal requirements more conservative
because (with the proposed arrangement) the Completion Time for
Condition A (and possibly Condition B) would start when the initial
inoperability occurred rather than (with literal interpretation of
the existing arrangement) when Condition A (or B) was entered after
the partial restoration. . . .
As described above, the proposed change corrects the structure
of the LCO to assure its correct application. There is no change in
intent or in the way the LCO is actually applied. The literal (and
unintended) interpretation of the existing LCO structure could,
under some circumstances, provide longer than intended Completion
Times for restoration of operability. The proposed change only
clarifies the requirements of the LCO Required Actions. Since the
proposed change affects neither the LCO intent nor its application,
the proposed change will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
B. [The proposed changes would not] create the possibility of a
new or different kind of accident from any previously evaluated.
As described above, the proposed change corrects the structure
of the LCO to assure its correct application. There is no change in
intent or in the way the LCO is actually applied. The proposed
changes would not result in any physical alterations to the plant
configuration, no new equipment is added, no equipment interfaces
are modified, no changes to any equipment's function or the method
of operating the equipment are being made. As the proposed changes
would not change the design, configuration or operation of the
plant, no new or different kinds of accident modes are created.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
C. [The proposed changes would not] involve a significant
reduction in a margin of safety.
As described above, the proposed change corrects the structure
of the LCO to assure its correct application. The proposed changes
are consistent with the intent of the changes made to the STS by
TSTF 325. There is no change in intent or in the way the LCO is
actually applied. 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: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: Claudia M. Craig.
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of amendment request: December 7, 2000.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) in accordance with changes to
the ``Standard Technical Specifications, Combustion Engineering
Plants,'' NUREG 1432, Revision 1 (STS), made by the Nuclear Energy
Institute Technical Specifications Task Force (TSTF) change number 325,
``Changes To Structure Of [Emergency Core Cooling System] ECCS--
Operating [Limiting Condition for Operation] LCO.'' Specifically,
Palisades LCO 3.5.2, ``ECCS--Operating,'' would be changed as follows:
(1) the second part of the Condition B description, ``At least 100% of
the required ECCS flow available,'' would be deleted; (2) the wording
of Condition C would be revised to limit its application to Conditions
A or B; and (3) the wording that would be removed from Condition B
would be made into a new condition, Condition D, which would read:
``Less than 100% of the required ECCS flow available.'' Required Action
D.1, ``Enter LCO 3.0.3,'' and its completion time, ``Immediately,''
would also be added. The licensee also forwarded related changes to the
TS Bases.
Additional changes requested in the licensee's application dated
December 7, 2000, are based upon other TSTFs and are addressed by
separate Federal Register notices.
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 change is proposed which emulates changes made to Standard
Technical Specifications, Combustion Engineering Plants, NUREG 1432,
Rev. 1 (STS) by TSTF 325. The structure of LCO 3.5.2, ECCS--
Operating, has been rearranged to maintain Condition B in effect if
failures should occur which reduce available flow to less than 100%
of the required flow (that flow assumed in the accident analyses).
The resulting requirements are those intended when the LCO was
initially constructed and represent the way the LCO Conditions are
being applied. Therefore there is no change in intent or application
of the LCO. In the case where inoperable ECCS train components
reduce available flow below that required, and a subsequent partial
restoration is made to provide 100% of the required flow, the
proposed change makes the literal requirements more conservative
because (with the proposed arrangement) the Completion Time for
Condition B would start when the initial inoperability occurred
rather than (with literal interpretation of the existing
arrangement) when Condition B was entered after the partial
restoration. * * *
A. [The proposed changes would not] involve a significant
increase in the probability or consequences of an accident
previously evaluated.
As described above, the proposed change corrects the structure
of the LCO to assure its correct application. There is no change in
intent or in the way the LCO is actually applied. The literal (and
unintended) interpretation of the existing LCO structure could,
under some circumstances, provide longer than intended Completion
Times for restoration of operability. The proposed change only
clarifies the requirements of the LCO Required Actions. Since the
proposed change affects neither the LCO intent nor its application,
the proposed change will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
B. [The proposed changes would not] create the possibility of a
new or different kind of accident from any previously evaluated.
As described above, the proposed change corrects the structure
of the LCO to assure its correct application. There is no change in
intent or in the way the LCO is actually applied. The proposed
changes would not
[[Page 7676]]
result in any physical alterations to the plant configuration, no
new equipment is added, no equipment interfaces are modified, and no
changes to any equipment's function or the method of operating the
equipment are being made. As the proposed changes would not change
the design, configuration or operation of the plant, no new or
different kinds of accident modes are created. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any previously evaluated.
C. [The proposed changes would not] involve a significant
reduction in a margin of safety.
As described above, the proposed change corrects the structure
of the LCO to assure its correct application. The proposed change is
consistent with the requirements of the STS. There is no change in
intent or in the way the LCO is actually applied. 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: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: Claudia M. Craig.
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of amendment request: December 7, 2000 (this application
supercedes an amendment request dated July 28, 2000).
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to allow Type B and C
containment leak rate testing to be performed in accordance with 10 CFR
part 50, appendix J, option B. Conversion to Option B affects TS 5.5.14
and Surveillance Requirements (SRs) SR 3.6.1.1, SR 3.6.1.3, and SR
3.6.2.1. The proposed amendment also revises the SR 3.6.2.2 frequency
for containment air lock door interlock testing from 18 months to 24
months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
* * * Four groups of changes have been proposed:
First, changes are proposed to allow Type B and C containment
leak rate testing to be performed in accordance with 10 CFR 50,
appendix J, Option B.
Second, exceptions are proposed to the Option B testing
methodology for containment air lock door seals.
Third, an exception is proposed to the Option B testing
frequency for small diameter containment purge valves.
Fourth, the frequency for the containment air lock door
interlock testing has been extended from 18 months to 24 months.
The following evaluation supports the finding that operation of
the facility in accordance with the proposed changes would not:
a. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
All four groups of proposed changes deal exclusively with
testing of features related to containment isolation. The changes
only affect testing frequency and methodology. The proposed testing
methodologies are acceptable under the existing Technical
Specifications. None of the devices involved are assumed as an
initiator of any accident previously evaluated. Therefore, operation
of the facility in accordance with the proposed changes would not
involve a significant increase in the probability of an accident.
1. The first group of proposed changes is based on the model
Technical Specifications approved by the NRC staff in TSTF
[Technical Specification Task Force] 52, Rev. 3. Test intervals will
be established based on performance history of the components
tested. The frequency of testing the containment penetrations and
containment isolation valves will be extended in accordance with
program requirements and 10 CFR 50, appendix J, Option B, with
reference to Regulatory Guide 1.163, and NEI [Nuclear Energy
Institute] 94-01, Rev 0. The change in risk resulting from the
proposed changes was evaluated by the NRC in the rule making process
for implementing the Option B requirements and are characterized in
NUREG-1493. For Type B and C tests the NRC concluded that the
extension of test intervals as allowed by Option B would lead to
only minor increases in potential offsite dose consequences. These
increases are offset by the expected decrease in worker dose
received during Type A, B, and C testing, and were found to be
acceptable. Therefore, operation of the facility in accordance with
the first group [of] proposed changes will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The second group of proposed changes would allow air lock
door seal leak rate testing to be performed by a seal contact check
(for the Emergency Escape Air Lock) or by pressurizing between the
door seals at a pressure [greater than or equal to] 10 psig (for the
Personnel Air Lock) following door seal contact adjustments. Both
proposed alternative testing methods are allowed by existing
Technical Specifications (while testing under Option A) and both
will result in a continuation of the currently successful testing
practice which has provided a high degree of confidence in door seal
performance. Plant operating history has shown that air lock door
seals which have been successfully tested in accordance with the
proposed methodology have passed subsequent full pressure air lock
leakage tests in virtually every case.
Since the proposed methodology has been demonstrated to
successfully detect leaking door seals, the continued use of that
methodology for testing under the requirements of Option B will not
cause an increase in the probability of a leaking air lock door seal
going undetected. Also, since there will be no increase in the rate
of occurance [sic] of undetected leakage due to the continued
utilization of current practices under Option B, operation of the
facility in accordance with the second group of proposed changes
will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
3. The third proposed change allows the testing frequency for
the Containment 4-inch purge exhaust, 8-inch purge exhaust and 12-
inch air room supply valves to be consistent with other 10 CFR 50,
appendix J, Option B, Type C test intervals and is supported by
Palisades design, historical test results and other required
testing. This would allow the test interval to be extended to a
maximum of 60 months from the 30 month interval allowed without this
exception.
The change in risk resulting from the third proposed change is
essentially the same as that evaluated by the NRC in the rule making
process for implementing the Option B Type C testing requirements,
which are characterized in NUREG-1493. As discussed under change 1,
above, the NRC concluded that the extension of test intervals as
allowed by Option B for Type C testing would lead to only minor
increases in potential offsite dose consequences. These increases
were found to be acceptable. The third proposed change applies this
longer interval to moderate diameter valves in the containment purge
system. That longer interval would apply to these valves, without
the proposed exception, if they were installed as containment
isolation valves in a different system. Furthermore, the 8-inch and
12-inch valves are effectively leak rate tested on a 184 day
frequency as part of their required closure verification. Therefore,
the proposed changes will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
4. The fourth proposed change only extends the frequency for
containment air lock door interlock testing. The proposed change
will not affect any parameters or conditions that contribute to the
mitigation of previously evaluated accidents. Therefore, operation
of the facility in accordance with the fourth proposed change would
not involve a significant increase in the consequences of an
accident previously evaluated.
b. Create the possibility of a new or different kind of accident
from any previously evaluated.
All four groups of proposed changes deal exclusively with
testing of features related to containment isolation. The changes
only affect testing frequency and methodology. The proposed testing
methodologies are acceptable under the existing Technical
Specifications. The proposed changes would not result in any
physical alterations to the
[[Page 7677]]
plant configuration, no new equipment is added, no equipment
interfaces are modified, no changes to any equipment's function or
the method of operating the equipment are being made. As the
proposed changes would not change the design, configuration or
operation of the plant, they would not cause the containment leak
rate testing to become an accident initiator. No new or different
kinds of accident modes are created. Therefore, the proposed changes
do not create the possibility of a new or different kind of accident
from any previously evaluated.
c. Involve a significant reduction in the margin of safety.
All four groups of proposed changes deal exclusively with
testing of features related to containment isolation. The changes
only affect testing frequency and methodology. The proposed testing
methodologies are acceptable under the existing Technical
Specifications. None of the devices involved are assumed as an
initiator of any accident previously evaluated. The proposed changes
only affect the methodology and frequency of Type B and C testing.
The methods for performing the tests are not changed from those
specified in existing Technical Specifications. The proposed
performance based approach, provided by using Option B to 10 CFR 50,
Appendix J, would continue to ensure that the containment leakage
rates would not exceed the maximum allowable leakage rates defined
in the Technical Specifications and assumed in the accident
analysis. 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: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: Claudia M. Craig.
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of amendment request: December 7, 2000.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) regarding the Limiting
Conditions for Operation (LCO) for the containment cooling systems (LCO
3.6.6), the component cooling water system (LCO 3.7.7), and the service
water system (LCO 3.7.8) to be similar to changes to the ``Standard
Technical Specifications, Combustion Engineering Plants,'' NUREG 1432,
Revision 1 (STS), made by the Nuclear Energy Institute Technical
Specifications Task Force (TSTF) change number 325, ``Changes To
Structure Of [Emergency Core Cooling System] ECCS--Operating LCO.''
Palisades LCO 3.6.6, ``Containment Cooling Systems,'' would be
changed as follows: (1) The second part of the Condition A description,
``AND--At least 100% of the required post accident containment cooling
capability available,'' would be deleted; (2) the wording of Condition
B would be revised to limit its application to Condition A; and (3) the
wording removed from Condition A would be made into a new condition,
Condition C, which would read: ``Less than 100% of the required post-
accident containment cooling capability available.'' Required Action
C.1, ``Enter LCO 3.0.3,'' and its completion time, ``Immediately,''
would also be added. The licensee also forwarded related changes to the
TS Bases.
Palisades LCO 3.7.7, ``Component Cooling Water [CCW] System,''
would be changed as follows: (1) The second part of the Condition A
description, ``AND--At least 100% of the required CCW post accident
capability available,'' would be deleted; (2) the wording of Condition
B would be revised to limit its application to Condition A; and (3) the
wording removed from Condition A would be made into a new condition,
Condition C, which would read: ``Less than 100% of the required post-
accident CCW capability available.'' Required Action C.1, ``Enter LCO
3.0.3,'' and its completion time, ``Immediately,'' would also be added.
The licensee also forwarded related changes to the TS Bases.
Palisades LCO 3.7.8, ``Service Water System [SWS],'' would be
changed as follows: (1) The second part of the Condition A description,
``AND--At least 100% of the required post accident SWS capability
available,'' would be deleted; (2) the wording of Condition B would be
revised to limit its application to Condition A; and (3) the wording
removed from Condition A would be made into a new condition, Condition
C, which would read: ``Less than 100% of the required post-accident SWS
capability available.'' Required Action C.1, ``Enter LCO 3.0.3,'' and
its completion time, ``Immediately,'' would also be added. The licensee
also forwarded related changes to the TS Bases.
Additional changes requested in the licensee's application dated
December 7, 2000, are based upon other TSTFs and are addressed by
separate Federal Register notices.
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:
Changes are proposed for three Palisades LCOs structured like
LCO 3.5.2 which emulate changes made to Standard Technical
Specifications, Combustion Engineering Plants, NUREG 1432, Rev.1
(STS) by TSTF 325. The structure of LCOs 3.6.6, 3.7.7, and 3.7.8 has
been rearranged to maintain Condition A in effect if failures should
occur which reduce available flow to less than 100% of the required
cooling capability (that assumed in the accident analyses). The
resulting requirements are those intended when the LCOs were
initially constructed and represent the way the LCO Conditions are
being applied. Therefore there is no change in intent or application
of the LCOs. In the case where inoperable required components reduce
available cooling below that required, and a subsequent partial
restoration is made to provide 100% of the required cooling, the
proposed change makes the literal requirements more conservative
because (with the proposed arrangement) the Completion Time for
Condition A would start when the initial inoperability occurred
rather than (with literal interpretation of the existing
arrangement) when Condition A was entered after the partial
restoration. * * *
A. [The proposed changes would not] involve a significant
increase in the probability or consequences of an accident
previously evaluated.
As described above, the proposed changes correct the structure
of the subject LCOs to assure their correct application. There is no
change in intent or in the way the LCOs are actually applied. The
literal (and unintended) interpretation of the existing LCO
structure could, under some circumstances, provide longer than
intended Completion Times for restoration of operability. The
proposed changes only clarify the requirements of the LCO Required
Actions. Since the proposed changes affect neither the LCO intent
nor their application, the proposed changes will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
B. [The proposed changes would not] create the possibility of a
new or different kind of accident from any previously evaluated.
As described above, the proposed changes correct the structure
of LCOs 3.6.6, 3.7.7, and 3.7.8 to assure their correct application.
There is no change in intent or in the way the LCOs are actually
applied. The proposed changes would not result in any physical
alterations to the plant configuration, no new equipment is added,
no equipment interfaces are modified, and no changes to any
equipment's function or the method of operating the equipment are
being made. As the proposed changes would not change the design,
configuration or operation of the plant, no new or different kinds
of accident modes are created. Therefore, the proposed
[[Page 7678]]
changes do not create the possibility of a new or different kind of
accident from any previously evaluated.
C. [The proposed changes would not] involve a significant
reduction in a margin of safety
As described above, the proposed changes correct the structure
of the subject LCOs to assure their correct application. The
proposed changes are consistent with the changes made to the STS by
TSTF 325. There is no change in intent or in the way the LCOs are
actually applied. 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: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: Claudia M. Craig.
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of amendment request: December 7, 2000.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) in accordance with changes to
the ``Standard Technical Specifications, Combustion Engineering
Plants,'' NUREG 1432, Revision 1, made by the Nuclear Energy Institute
Technical Specifications Task Force (TSTF) change number 258, Revision
4. TSTF 258 addresses changes to various Administrative Controls TSs.
The licensee proposes the following four changes to the Palisades TSs:
(1) In section 5.2, ``Organization,'' Palisades TS Section 5.5.2e
would be revised by deleting the specific detail of working hour
limitations (i.e., administrative procedures are used to control
working hours).
(2) Also in Section 5.2, TS Section 5.5.2g would be revised by
deleting the title for the ``Shift Technical Advisor'' position and by
clarifying the requirements for that position.
(3) In TS Section 5.5.4, ``Radioactive Effluent Controls Program,''
sections 5.5.4b, 5.5.4e, and 5.5.4h would be revised to be consistent
with 10 CFR part 20.
(4) TS Section 5.7, ``High Radiation Area,'' would be revised to be
consistent with 10 CFR Part 20.1601(c) (i.e., the existing TS would be
completely replaced by Insert F from TSTF 258).
Additional changes requested in the licensee's application dated
December 7, 2000, are based upon other TSTFs and are addressed by
separate Federal Register notices.
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 would not] involve a significant
increase in the probability or consequences of an accident
previously evaluated.
All four proposed changes deal exclusively with Administrative
Controls. The changes only affect the details of controls placed on
the plant staff and their working conditions. The proposed controls
are consistent with the requirements approved for STS. None of the
controls involved are assumed to be associated with any initiator
of, or any mitigating equipment or mitigation actions for any
accident previously evaluated. Therefore, the proposed changes will
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
B. [The proposed changes would not] create the possibility of a
new or different kind of accident from any previously evaluated.
All four proposed changes deal exclusively with Administrative
Controls. The changes only affect the details of controls placed on
the plant staff and their working conditions. The proposed controls
are consistent with the requirements approved for STS. The proposed
changes would not result in any physical alterations to the plant
configuration, no new equipment is added, no equipment interfaces
are modified, no changes to any equipment's function or the method
of operating the equipment are being made. As the proposed changes
would not change the design, configuration or operation of the
plant, no new or different kinds of accident modes are created.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
C. [The proposed changes would not] involve a significant
reduction in a margin of safety.
All four proposed changes deal exclusively with Administrative
Controls. The changes only affect the details of controls placed on
the plant staff and their working conditions. The proposed controls
are consistent with the requirements approved for STS. None of the
controls involved are assumed to be associated with any initiator
of, or any mitigating equipment or mitigation actions for any
accident previously evaluated. 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: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: Claudia M. Craig.
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of amendment request: December 7, 2000.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) in accordance with changes to
the ``Standard Technical Specifications, Combustion Engineering
Plants,'' NUREG 1432, Revision 1 (STS), made by the Nuclear Energy
Institute Technical Specifications Task Force (TSTF) change number 287,
``Allowances For Breach Of The Control Room Envelope,'' Revision 5.
Specifically, a note would be added modifying TS section 3.7.10,
``Control Room Ventilation (CRV) Filtration,'' to allow the control
room boundary to be opened intermittently under administrative control,
and a new condition (Condition B) would be added to the Action table
for TS section 3.7.10 to allow 24 hours to restore an inoperable
control room boundary. A required Action (B.1) would also be added
requiring certain preplanned actions to be initiated immediately upon
discovery that the containment envelope is inoperable. The subsequent
conditions and required actions would be renumbered accordingly and
supporting editorial changes would be made to the descriptions for
Conditions B and E (to be renumbered as Conditions C and F). The
licensee also forwarded related changes to the Bases for TS section
3.7.10.
Additionally, a correction would be made to the Action table for TS
Section 3.7.10 by restoring Required Action D.2 (to be renumbered to
E.2), which was inadvertently omitted during the prior issuance of the
Palisades Improved TSs by Amendment No. 189.
Additional changes requested in the licensee's application dated
December 7, 2000, are based upon other TSTFs and are addressed by
separate Federal Register notices.
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:
[[Page 7679]]
A. [The proposed changes would not] involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes deal exclusively with allowances to
temporarily deviate from the [Limiting Condition for Operation] LCO
3.7.10 requirement (established by [Surveillance Requirement] SR
3.7.10.4) for the control room boundary to be sufficiently air tight
to maintain 0.125 inches of water differential when the ventilation
system is in the emergency mode of operation. The proposed controls
are consistent with the requirements approved for STS. None of the
controls involved are assumed to be associated with any assumed
initiator of any accident previously evaluated.
The proposed changes do allow temporary (up to 24 hours)
relaxation of controls put in place to protect the operators from
accidental releases of particulate radioactive materials. The
utilization of this temporary allowance is expected to be
infrequent, and the controls required when this allowance is
utilized maintain the intended radiological protection for the
operators in the control room areas. Since the protection of the
operators in the control room areas will be provided by alternate
means during the exercising of these allowances, there will be no
effect on their perceived abilities to mitigate the consequences of
an accident.
Therefore, operation of the facility in accordance with the
proposed changes will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
B. [The proposed changes would not] create the possibility of a
new or different kind of accident from any previously evaluated.
The proposed changes deal exclusively with an allowance to
temporarily provide radiological protection within the control room
boundary by alternative means. The proposed controls are consistent
with the requirements approved for STS. The proposed changes would
not result in any physical alterations to the operating plant
systems, no new equipment is added, no equipment interfaces are
modified, no changes to any equipment's function or the method of
operating the power generation or accident mitigating equipment are
being made. As the proposed changes would not change the design,
configuration or operation of the plant, no new or different kinds
of accident modes are created. Therefore, the proposed changes do
not create the possibility of a new or different kind of accident
from any previously evaluated.
C. [The proposed changes would not] involve a significant
reduction in a margin of safety.
The proposed changes deal exclusively with an allowance to
temporarily provide radiological protection within the control room
boundary by alternative means. The proposed controls are consistent
with the requirements approved for STS. None of the controls
involved are assumed to be associated with any initiator of, or any
mitigating equipment or mitigation actions for any accident
previously evaluated. 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: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: Claudia M. Craig.
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of amendment request: December 8, 2000.
Description of amendment request: The proposed amendment would
delete requirements from the Technical Specifications (TSs) (and, as
applicable, other elements of the licensing bases) to maintain a post
accident sampling system (PASS). Licensees were generally required to
implement PASS upgrades as described in NUREG-0737, ``Clarification of
TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power
Plants to Assess Plant and Environs Conditions During and Following an
Accident.'' Implementation of these upgrades was an outcome of the
lessons learned from the accident that occurred at TMI, Unit 2.
Requirements related to PASS were imposed by Order for many facilities
and were added to or included in the TSs for nuclear power reactors
currently licensed to operate. Lessons learned and improvements
implemented over the last 20 years have shown that the information
obtained from PASS can be readily obtained through other means or is of
little use in the assessment and mitigation of accident conditions.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 11, 2000 (65 FR 49271), on possible
amendments to eliminate PASS, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on October 31, 2000 (65 FR 65018). The licensee affirmed the
applicability of the following NSHC determination in its application
dated December 8, 2000.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase
[[Page 7680]]
in the consequences of any accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Previously
Evaluated.
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radionuclides
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: Claudia M. Craig.
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
Date of amendment request: October 24, 2000.
Description of amendment request: Entergy Operations, Inc. is
proposing that the Grand Gulf Nuclear Station (GGNS) Operating License
be amended to revise the GGNS Technical Specifications (TSs), which
govern the lube oil inventories for the Division I, II, and III
Emergency Diesel Generators (EDGs). The change would increase the lube
oil inventories specified in TS 3.8.3 to ensure continued operation of
the EDGs under post-accident conditions, and provide additional margin
in lube oil consumption calculations. The TS change would account for
potential increases in EDG lube oil consumption rates which exceed the
nominal consumption rates originally used to determine EDG lube oil
requirements to support seven days of EDG operation at rated load
conditions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
The purpose of the emergency diesel generators is to mitigate
the consequences of analyzed accidents. Emergency Diesel Engine
inoperability or loss of capability has no effect on the probability
of any analyzed accident. The reason for this change is to provide
added assurance that the engines perform per the design requirements
and therefore the consequences of an accident previously evaluated
are not increased.
The purpose of the requested change is to regain margin in the
lube oil consumption calculations, such that, if increases in
consumption should occur in the future, Technical Specifications
requirements will still ensure operability of the Diesel Generators.
Design Engineering has basically taken the vendor's specified
consumption rate and doubled that value to ensure that the newly
calculated inventory limit will bound any potential consumption rate
increases.
Current calculations using as found consumption rates have shown
that the limiting sump volume is on Division III engines and that
there is minimal margin left between the actual volume and the
calculated volume needed. Therefore, there is a need for an external
dedicated storage skid, which is the only physical change to the
plant necessary to support this change request. The current
licensing basis recognizes that make-up oil may be required at some
point during a design basis event. The current Bases for Technical
Specification 3.8.3 LCO provides this recognition.
Given the stated purpose and no need for changes to installed
plant structures, (other than addition of a new Div[ision] III lube
oil storage skid) systems, or components there will be no
significant changes to the operation of the facility. Therefore,
this change does not involve a significant increase in the
probability or consequences of any accident previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
The purpose of the emergency diesel generators is to mitigate
the consequences of analyzed accidents; the engines are not accident
initiating. Emergency Diesel Engine inoperability or loss of
capability cannot create the possibility of a new or different kind
of accident from any accident previously evaluated. The reason for
this change is to provide added assurance that the engines perform
per the design requirements.
The Diesel Engine Lubricating System (DELS) design and operation
is unaffected by his change. Recognizing the need for having a make-
up inventory and staging a volume readily accessible to the operator
will enhance the operator's ability to maintain DG [Diesel
Generator] operable. Design Engineering has performed appropriate
fire hazards reviews and seismic II/l reviews to assure compliance
with current design requirements.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
The current licensing basis requires that the DELS provide seven
days of Diesel Generator operation under specified load conditions.
This basis was substantiated via calculation using vendor supplied
consumption rates of 1.21 (Div[ision] I and II) and 0.6 (Div[ision]
Ill) gallons per hour. The current basis recognizes that make-up oil
may be required at some point during a design basis event. To ensure
this basis is valid for future operations, Design Engineering has
recalculated the required inventories based on a more conservative
consumption rate. This change will ensure that sufficient lube oil
is readily available to support the extended run times under post
accident conditions. Therefore, this change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334,
Beaver Valley Power Station, Unit No. 1, Shippingport, Pennsylvania
Date of amendment request: December 21, 2000.
Description of amendment request: The amendment request proposes to
delete the Steam/Feedwater Flow Mismatch coincident with Low Steam
[[Page 7681]]
Generator (SG) Water Level reactor trip from the technical
specifications. The Steam/Feedwater Flow Mismatch coincident with Low
SG Water Level reactor trip was included in the Unit 1 design in order
to meet regulatory requirements regarding potentially adverse control
and protection system interactions. The amendment request proposes to
take credit for the SG Level Median Selector Switch (MSS) installed in
1997 to meet these requirements. The MSS eliminates the potential for
an adverse control and protection system interaction and, therefore,
eliminates the design requirement for the Steam/Feedwater Flow Mismatch
and Low SG Level reactor trip. Appropriate changes to the Bases are
also included in the amendment request.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The initiating conditions and assumptions for accidents
described in the Updated Final Safety Analys[i]s Report remain as
previously analyzed. The proposed change does not introduce a new
accident initiator nor does it introduce changes to any existing
accident initiators or scenarios described in the Updated Final
Safety Analys[i]s Report. The Steam/Feedwater Flow Mismatch and Low
Steam Generator Water Level reactor trip is not credited for
accident mitigation in any accident analyses described in the
Updated Final Safety Analys[i]s Report. The Steam/Feedwater Flow
Mismatch and Low Steam Generator Water Level trip was designed to
meet the control and protection systems interaction criteria of the
Institute of Electric and Electronic Engineers Standard 279. The
Median Selector Switch prevents adverse control and protection
system interaction such that it replaces the need for the Steam/
Feedwater Flow Mismatch and Low Steam Generator Water Level reactor
trip and satisfies the Institute of Electric and Electronic
Engineers Standard 279 requirements. As such, the affected control
and protection systems will continue to perform their required
functions without adverse interaction and the capability to shut
down the reactor when required on Low-Low Steam Generator water
level to mitigate an accident previously evaluated is unaffected.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The substitution of the Median Selector Switch for the Steam/
Feedwater Flow Mismatch and Low Steam Generator Water Level trip
will not introduce any new failure modes to the required protection
functions. The Median Selector Switch only interacts with the
feedwater control system and the Steam Generator Water Level Low-Low
protection function is not affected by this change. Isolation
devices in the Median Selector Switch circuitry ensure that the
Steam Generator Water Level Low-Low protection function is not
affected. The Median Selector Switch is designed to reduce the
frequency of system failures through utilization of highly reliable
components in a design that relies on a minimum of additional
equipment. Components utilized in the Median Selector Switch are of
a quality consistent with low failure rates and minimum maintenance
requirements, and conform to protection system requirements.
Furthermore, the design provides the capability for complete unit
testing that provides unambiguous determination of credible system
failures. It is through these features that the overall design of
the Median Selector Switch minimizes the occurrence of undetected
failures that may exist between test intervals. Additionally, the
reliability of the Median Selector Switch has been shown by Unit 2
operating experience to be acceptable.
3. Does the change involve a significant reduction in a margin
of safety?
The margin of safety depends on the maintenance of specific
operating parameters and systems within design requirements and
safety analysis assumptions.
The proposed amendment does not involve revisions to any safety
limits or safety system setting that would adversely impact plant
safety. The proposed amendment does not alter the functional
capabilities assumed in a safety analysis for any system, structure,
or component important to the mitigation and control of design bases
accident conditions within the facility. Nor does this amendment
revise any parameters or operating restrictions that are assumptions
of a design basis accident. In addition, the proposed amendment does
not affect the ability of safety systems to ensure that the facility
can be placed and maintained in a shutdown condition for extended
periods of time.
The ability of the Steam Generator Water Level Low-Low reactor
trip function credited in the safety analysis to protect against a
sudden loss of heat sink event is not affected by the proposed
change. Since the Steam Generator Low-Low Level trip provides
complete protection for all accident transients that result in low
steam generator level, eliminating the Steam/Feedwater Flow Mismatch
and Low Steam Generator Water Level trip will not change any safety
analysis conclusion for any analyzed accident described in the
Updated Final Safety Analys[i]s Report.
The Median Selector Switch prevents adverse control and
protection system interaction such that it replaces the need for the
Steam/Feedwater Flow Mismatch and low Steam Generator Water Level
reactor trip and satisfies the Institute of Electric and Electronic
Engineers Standard 279 requirements. The proposed change will
enhance safe operation since the Steam/Feedwater Flow Mismatch and
Low Steam Generator Water Level trip function removal decreases the
challenges to the plant safety systems, decreases the plant
surveillance/maintenance activity, and reduces the plant complexity;
all resulting in a reduction in the potential for unnecessary plant
transients.
The technical specifications continue to assure the applicable
operating parameters and systems are maintained within the design
requirements and safety analysis assumptions. Therefore, the
elimination of this trip function will not result in a significant
reduction in the margin of safety as defined in the Updated Final
Safety Analys[i]s Report or technical specifications.
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 O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Marsha Gamberoni.
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit 1, Berrien County, Michigan
Date of amendment request: January 2, 2001.
Description of amendment requests: The proposed amendment would
revise Technical Specifications (TS) 3/4.6.2.2.a for the Unit 1 spray
additive tank to require a contained volume between 4000 and 4600
gallons of between 30 and 34 percent by weight sodium hydroxide (NaOH)
solution. In addition, the proposed amendment would make four types of
format changes to the revised Unit 1 page:
1. Reformat the header to include numbered first and second tier TS
section titles and a full-width single line to separate the header
section titles from the page text.
2. Reformat the footer to include ``COOK NUCLEAR PLANT--UNIT1'' on
the left side of the page, ``Page (page number)'' center page,
``AMENDMENT (past amendment numbers, with strikethrough, and ending
with the current amendment number)'' on the right side, and a full-
width single line to separate the footer from the page text.
3. Delete the double lines under ``LIMITING CONDITION FOR
OPERATION'' and ``SURVEILLANCE REQUIREMENTS.''
4. Fully justify the text and change the font.
Basis for proposed no significant hazards consideration
determination:
[[Page 7682]]
As required by 10 CFR 50.91(a), the licensee has provided its analysis
of the issue of no significant hazards consideration, which is
presented below:
1. Does the change involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated?
Adding a maximum limit for the allowed contained volume and
[sodium hydroxide] NaOH concentration for the spray additive tank
does not increase the probability of occurrence of any accident. The
spray additive system cannot initiate any previously analyzed
accident. The proposed changes ensure that the spray additive system
and the associated containment spray system can perform the accident
mitigation functions required during a [loss-of-coolant accident]
LOCA or [main steam line break] MSLB event. This action does not
affect the initiating frequency of a LOCA or MSLB event. Therefore,
the proposed changes do not increase the probability of an accident
previously evaluated.
The accidents previously evaluated in Chapter 14 of the Updated
Final Safety Analysis Report that are possibly affected by operation
of the spray additive system are a loss-of-coolant accident (LOCA)
and a main steam line break (MSLB). These postulated accidents are
expected to result in a containment spray signal, which then results
in the automatic starting of the containment spray pumps and the
opening of the valves associated with the spray additive system. The
spray additive system adds NaOH to the containment spray water being
supplied from the refueling water storage tank (RWST) to adjust the
pH of the containment spray and containment recirculation sump
solutions.
Following a LOCA, the containment spray water becomes mixed in
the containment recirculation sump with ice melt from the ice
condenser, reactor coolant from the reactor coolant system (RCS),
water being injected to the RCS from the safety injection
accumulators, and water being injected to the RCS from the RWST by
the emergency core cooling system. Following a MSLB, the containment
spray water becomes mixed in the containment recirculation sump with
ice melt from the ice condenser and the secondary coolant released
from the ruptured steam line.
The existing minimum and proposed maximum limits for the
contained volume and NaOH concentration for the spray additive tank
ensure a pH value of between 7.6 and 9.5 for the solution
recirculated within containment after a LOCA. This pH band minimizes
the evolution of iodine from the containment recirculation sump, and
minimizes the effect of chloride and caustic stress corrosion on
mechanical systems and components. An increase in pH value to at
least 7.0 in the containment recirculation sump during the
recirculation phase following a LOCA is consistent with the iodine
retention assumptions of the accident analyses. Therefore, the
consequences of a LOCA remain unchanged by the proposed changes. For
a MSLB, there is no increase in consequences since the containment
spray system and containment recirculation sump are not credited for
removal and retention of fission products from the containment
atmosphere.
The analyses for determining hydrogen generation following a
large break LOCA assume a specific pH time-dependent profile for the
containment spray and containment recirculation sump solutions. The
existing minimum and proposed maximum limits for the contained
volume and NaOH concentration for the spray additive tank do not
result in an increase in the previously predicted hydrogen
generation rates. Therefore, the current hydrogen generation
analyses remain bounding.
For both LOCA and MSLB events, the existing minimum and proposed
maximum limits for the contained volume and NaOH concentration for
the spray additive tank ensure that the pH of the containment spray
solution is within the bounds used in evaluations for environmental
qualification of required equipment.
Therefore, the proposed changes cannot increase the probability
of occurrence or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Adding a maximum allowed contained volume and NaOH concentration
for the spray additive tank does not create the possibility of an
accident of a new or different type than any previously evaluated.
The proposed changes ensure that the spray additive system, and the
associated containment spray system, can perform the required
accident mitigation functions during a LOCA or MSLB event. There are
no other types of accidents that can be postulated that would
require the use of the spray additive system or the associated
containment spray system for mitigation. The proposed changes do not
introduce any new association between the spray additive system and
any radioactive system, including the RCS. Therefore, emergency
operation of the spray additive system, or postulated failures of
the spray additive system, cannot initiate any type of accident.
Therefore, the proposed changes do not increase the possibility
of a new or different kind of accident than previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed limits on maximum allowed contained volume and NaOH
concentration for the spray additive tank ensure that the original
margin of safety is maintained by ensuring acceptable pH control
following a LOCA or MSLB event. Therefore, the proposed changes
ensure that the margin of safety is maintained by limiting the
maximum pH of the containment spray and containment recirculation
sump solutions following a LOCA or MSLB event.
Therefore, the proposed changes do 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 requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: Claudia M. Craig.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: October 24, 2000.
Description of amendment requests: The proposed amendments would
approve an unreviewed safety question allowing the use of new
methodology to calculate the transient response to steam generator tube
ruptures.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated?
The proposed change, to adopt a new analytical method to
evaluate the effects of an [Steam Generator Tube Rupture] SGTR, does
not affect any accident initiators or precursors. As such, the
proposed change does not increase the probability of an accident.
The proposed change also does not affect the ability of operators to
mitigate the consequences of an accident. The proposed change does
not impact the design of the affected plant systems such that
previously analyzed systems, structures, and components (SSCs) would
now be more likely to fail. The changes will not modify plant
systems to reduce their design capability during normal operating
and accident conditions. The use of the WCAP-10698-P-A methodology
to more accurately calculate the flow from the reactor coolant
system (RCS) to the SG secondary side following a postulated SGTR
does not affect the probability of any analyzed events. The use of
the WCAP-10698-P-A methodology does not affect SGTR initiators or
precursors. Therefore, incorporating the new methodology does not
affect equipment malfunction probability, nor does it affect or
create new accident initiators or precursors. Thus, there will be no
reduction in the capability of those SSCs in limiting the
consequences of previously evaluated accidents.
Additionally, the present methodology for calculating the
radiological consequences of a postulated SGTR is conservative when
compared with results from the new methodology. As such, the
existing licensing basis radiological consequence calculations will
be retained. Thus, no additional radiological source terms are
generated, and the consequences of an accident previously
[[Page 7683]]
evaluated in the [updated final safety analysis report] UFSAR will
not be increased. The use of this WCAP methodology and associated
computer code for break flow modeling more accurately calculates the
plant response to an SGTR event. The improved accuracy of the new
methodology provides valuable information related to the analysis of
operator actions and the associated timing. Such accurate transient
response information enables enhancements to be made to the
emergency operating procedures (EOPs).
Therefore, the proposed changes cannot increase the consequences
or probability of occurrence 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 change does not impact the design of affected plant
systems, involve a physical alteration to the systems, or change to
the way in which systems are currently operated, such that
previously unanalyzed SGTRs would now occur. The change to
incorporate the WCAP-10698-P-A methodology does not introduce any
new malfunctions; it calculates more accurately the flow from the
RCS to the SG secondary side following a postulated SGTR to
determine the time available for operator actions to prevent
overfilling the affected SG.
Thus, use of the WCAP-10698-P-A methodology does not affect or
create new accident initiators or precursors or create the
possibility of a new or different kind of accident.
Therefore, the proposed changes do not increase the possibility
of a new or different kind of accident than previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The approval of the license amendment will not result in any
modifications to affected plant systems that would reduce their
design capabilities during normal operating and accident conditions.
By using the WCAP-10698-P-A methodology, a more accurate SGTR
response is calculated. The improved understanding of the transient
response enables enhancements to the EOPs, which provide further
assurance that SSCs required for accident mitigation are protected.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
In summary, based upon the above evaluation, I&M has concluded
that these changes involve no significant hazards.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: Claudia M. Craig.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: December 18, 2000.
Description of amendment request: The proposed amendment deletes
requirements from the Technical Specifications (and, as applicable,
other elements of the licensing bases) to maintain a Post Accident
Sampling System (PASS). Licensees were generally required to implement
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI, Unit 2. Requirements related to
PASS were imposed by Order for many facilities and were added to or
included in the technical specifications (TS) for nuclear power
reactors currently licensed to operate. Lessons learned and
improvements implemented over the last 20 years have shown that the
information obtained from PASS can be readily obtained through other
means or is of little use in the assessment and mitigation of accident
conditions.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 11, 2000 (65 FR 49271) on possible
amendments to eliminate PASS, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on October 31, 2000 (65 FR 65018). The licensee affirmed the
applicability of the following NSHC determination in its application
dated December 18, 2000.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result of
the TMI-2 accident. The specific intent of the PASS was to provide a
system that has the capability to obtain and analyze samples of plant
fluids containing potentially high levels of radioactivity, without
exceeding plant personnel radiation exposure limits. Analytical results
of these samples would be used largely for verification purposes in
aiding the plant staff in assessing the extent of core damage and
subsequent offsite radiological dose projections. The system was not
intended to and does not serve a function for preventing accidents and
its elimination would not affect the probability of accidents
previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual benefit
to post accident mitigation. Past experience has indicated that there
exists in-plant instrumentation and methodologies available in lieu of
a PASS for collecting and assimilating information needed to assess
core damage following an accident. Furthermore, the implementation of
Severe Accident Management Guidance (SAMG) emphasizes accident
management strategies based on in-plant instruments. These strategies
provide guidance to the plant staff for mitigation and recovery from a
severe accident. Based on current severe accident management strategies
and guidelines, it is determined that the PASS provides little benefit
to the plant staff in coping with an accident.
The regulatory requirements for the PASS can be eliminated without
degrading the plant emergency response. The emergency response, in this
sense, refers to the methodologies used in ascertaining the condition
of the reactor core, mitigating the consequences of an accident,
assessing and projecting offsite releases of radioactivity, and
establishing protective action recommendations to be communicated to
offsite authorities. The elimination of the PASS will not prevent an
accident management strategy that meets the initial intent of the post-
TMI-2 accident guidance through the use of the SAMGs, the emergency
plan (EP), the emergency operating procedures (EOP), and site survey
monitoring that support modification of emergency plan protective
action recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
[[Page 7684]]
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of
a New or Different Kind of Accident from any Previously Evaluated.
The elimination of PASS related requirements will not result in any
failure mode not previously analyzed. The PASS was intended to allow
for verification of the extent of reactor core damage and also to
provide an input to offsite dose projection calculations. The PASS is
not considered an accident precursor, nor does its existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within the
containment building.
Therefore, this change does not create the possibility of a new or
different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
The elimination of the PASS, in light of existing plant equipment,
instrumentation, procedures, and programs that provide effective
mitigation of and recovery from reactor accidents, results in a neutral
impact to the margin of safety. Methodologies that are not reliant on
PASS are designed to provide rapid assessment of current reactor core
conditions and the direction of degradation while effectively
responding to the event in order to mitigate the consequences of the
accident. The use of a PASS is redundant and does not provide quick
recognition of core events or rapid response to events in progress. The
intent of the requirements established as a result of the TMI-2
accident can be adequately met without reliance on a PASS.
Therefore, this change does not involve a significant reduction in
the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Section Chief: James W. Clifford
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of amendment request: July 31, 2000.
Description of amendment request: The proposed license amendment
will change the method used to determine the Fuel Centerline Melt
Linear Heat Rate Limit (FCMLHRL). The proposed change represents a
departure from the use of the fixed value of 21 kilowatts per foot for
the FCMLHRL, which is being used in the current operating cycle, to a
value that will be calculated on a cycle-by-cycle basis using the
Siemens Power Corporation (SPC) U.S. Nuclear Regulatory Commission
(NRC) approved methodology. Northeast Nuclear Energy Company (the
licensee) has evaluated this proposed method of calculating FCMLHRL
utilizing the criteria of 10 CFR 50.59. The licensee has determined
that this change involves an unreviewed safety question (USQ). The
licensee is requesting approval of the USQ.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
This license amendment request deals with changes in the
Millstone Unit No. 2 Final Safety Analysis Report (FSAR) due to
changing the method used to determine the FCMLHRL. The proposed
change represents a departure from the use of the fixed value of 21
kW/ft for the FCMLHRL, which is being used in the current cycle, to
a value that will be calculated on a cycle by cycle basis using the
SPC approved methodology. This methodology was reviewed and approved
by the Nuclear Regulatory Commission (NRC) and is documented in
Siemens Power Corporation (SPC) report XN-NF-82-06(P)(A).[ ] The
value of the FCMLHRL is verified for each reload, but does not
typically change significantly between cycles. This limit is
determined for a standard fuel rod. The current enrichment cutbacks
in the gadolinia bearing rods limit their relative power such that
the maximum FCMLHRL for a gadolinia bearing fuel rod will be
sufficiently below the standard fuel rods to prevent centerline
melt. In future applications of this methodology, the peak Linear
Heat Rates (LHR) calculated from transient analyses will be compared
to the FCMLHRL for the cycle. The Local Power Density (LPD) Limiting
Safety System Settings (LSSS) verification analysis for future
applications will use the cycle dependent FCMLHRL. Therefore, It can
be concluded that these FSAR changes are safe and that the cycle
specific calculated FCMLHRL has no impact on plant equipment
operation. Further more, the change in the method of determining the
FCMLHRL only impacts the analytical determination of failed fuel and
has no direct impact on the accident scenario. Accordingly, this
change cannot affect the likelihood of these events. Therefore, the
proposed changes will not increase the probability of occurrence of
accidents previously evaluated.
The change in the method of determining the FCMLHRL will
continue to conservatively estimate fuel failures. Since the
proposed FSAR changes will have no impact on the analysis of the
events, they cannot affect the likelihood or consequences of these
events. Therefore, the proposed FSAR changes will not increase the
consequences of accidents previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed FSAR changes will not alter the plant configuration
(no new or different type of equipment will be installed) or require
any new or unusual operator actions. The FSAR changes do not
introduce any new failure modes. Therefore, the changes will not
increase the probability of a new or different kind of accident from
any accidents previously evaluated.
3. Involve a significant reduction in a margin of safety.
The purpose of the proposed changes is to document a change in
the method used to determine FCMLHRL in the Millstone Unit No. 2
FSAR. The change in methodology may result in a FCMLHRL that is
higher than the previous limit of 21 kW/ft. Therefore, the proposed
changes may lead to a reduction of the margin of safety. However,
the proposed changes are safe because SPC has justified, using NRC
generically approved methodology, that with a higher value of the
FCMLHRL the fuel will not experience centerline melt. In other
words, a higher FCMLHRL may allow a higher fuel temperature but will
continue to protect fuel against centerline melt. Therefore, it can
be concluded that the FSAR changes are safe and do not significantly
reduce 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: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Section Chief: James W. Clifford.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: December 13, 2000.
Description of amendment request: The proposed amendment would
[[Page 7685]]
change License Condition 2.C.4 to conform to NRC Generic Letter (GL)
86-10, ``Implementation of Fire Protection Requirements.'' The proposed
amendment would also relocate the Fire Protection Program (FPP)
elements from the Technical Specifications (TSs) to the licensee-
controlled FPP, in accordance with GL 86-10 and GL 88-12, ``Removal of
Fire Protection Requirements from Technical Specifications.''
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 amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The requested changes are administrative in nature in that they
move fire protection requirements from the TS to the FPP and
associated implementing procedures following the guidance of NRC
Generic Letter (GL) 86-10 and GL 88-12. The requested changes will
not revise the requirements for fire protection equipment
operability, testing or inspections.
The proposed changes do not involve any change to the
configuration or method of operation of any plant equipment that is
used to mitigate the consequences of an accident, nor do they affect
any assumptions or conditions in any of the accident analyses. Since
the accident analyses remain bounding, their radiological
consequences are not adversely affected.
Therefore, the probability or consequences of an accident
previously evaluated are not affected.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
The requested changes are administrative in nature in that they
move fire protection requirements from the TS to the FPP and
associated implementing procedures following the guidance of GL 86-
10 and 88-12. The requested changes will not revise the requirements
for fire protection equipment operability, testing or inspections.
The proposed changes do not involve any change to the
configuration or method of operation of any plant equipment that is
used to mitigate the consequences of an accident, nor do they affect
any assumptions or conditions in any of the accident analyses.
Accordingly, no new failure modes have been defined for any plant
system or component important to safety nor has any new limiting
single failure been identified as a result of the proposed changes.
Therefore the possibility of a new or different kind of accident
from any accident previously evaluated is not created.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
The requested changes are administrative in nature in that they
move fire protection requirements from the TSs to the FPP and
associated implementing procedures following the guidance of GL 86-
10 and 88-12. The requested changes will not revise the requirements
for fire protection equipment operability, testing or inspections.
Future changes to the program will be reviewed in accordance with
the fire protection license condition to ensure that the ability to
achieve and maintain safe shutdown in the event of a fire are [sic]
not adversely affected.
Therefore, a significant reduction in the margin of safety is
not involved.
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: Jay E. Silberg, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Section Chief: Claudia M. Craig.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: December 13, 2000.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.8/4.8 to clarify the air ejector
offgas activity sample point and operability requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes clarify and more completely specify actions and
requirements with respect to main condenser offgas activity.
Compliance with applicable regulatory requirements will continue to
be maintained. The proposed changes do not alter the conditions or
assumptions in any of the previous accident analyses. Since the
previous accident analyses remain bounding, the radiological
consequences previously evaluated are not adversely affected by the
proposed changes.
Therefore, the probability or consequences of an accident
previously evaluated are not affected by any of the proposed
amendments.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not involve any change to the method of
operation of any plant equipment. Accordingly, no new failure modes
have been defined for any plant system or component important to
safety nor has any new limiting single failure been identified as a
result of the proposed changes. Also, there will be no change in
types or increase in the amounts of any effluents released offsite.
Therefore, the possibility of a new or different kind of
accident from any accident previously evaluated would not be
created.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
The proposed changes do not involve a significant reduction in a
margin of safety. The proposed changes clarify and more completely
specify actions and requirements with respect to main condenser
offgas activity. No changes in radioactivity release limits or dose
limits are proposed. The changes in actions to be taken if a limit
is not met provide an adequate means of ensuring that the health and
safety of the public are protected and that potential dose to the
public is below regulatory limits. The proposed changes do not
involve any actual change in the methodology used in the control of
radioactive effluents. The proposed changes also comply with the
guidance contained in the STS [standard technical specifications].
Therefore, a significant reduction in the margin of safety is
not involved.
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: Jay E. Silberg, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Section Chief: Claudia M. Craig.
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 amendment requests: December 6, 2000.
Description of amendment requests: The proposed license amendments
would revise Section 5.0, ``Administrative Controls,'' of the Diablo
Canyon Power Plant, Unit Nos. 1 and 2 Technical Specifications (TS) to
change the following management titles.
(1) TS 5.1.1 would be revised to replace the titles ``Vice
President, Diablo Canyon Operations and Plant Manager,'' and ``Plant
Manager,'' with the generic title ``plant manager.''
[[Page 7686]]
(2) TS 5.2.1.a, last sentence, would be revised to state: ``These
requirements, including the plant-specific titles of those personnel
fulfilling the responsibilities of the positions delineated in these
Technical Specifications, shall be documented in the FSAR [Final Safety
Analysis Report] Update.''
(3) TS 5.2.1.b, would be revised to replace the title ``Plant
Manager,'' with the generic title ``plant manager.''
(4) TS 5.2.1.c. would be revised to replace the title ``Senior Vice
President and General Manager--Nuclear Power Generation,'' with the
generic title ``specified corporate officer.''
(5) TS 5.2.2.d would be revised to replace the title ``Plant
Manager,'' with the generic title ``plant manager.''
(6) TS 5.2.2.e. would be revised to replace the title ``Operations
Director'' with the generic title ``operations manager.''
(7) TS 5.3.1 would be revised to replace the titles ``Radiation
Protection Director'' and ``Operations Director'' with the generic
titles ``radiation protection manager'' and ``operations manager,''
respectively.
(8) TS 5.5.1.b (second paragraph ``b'') would be revised to replace
the title ``Plant Manager,'' with the generic title ``plant manager.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
This License Amendment Request (LAR) proposes to revise
Technical Specification (TS) 5.0, ``Administrative Controls,'' to
replace specific management titles with lower case generic titles
consistent with Industry/Technical Specification Task force (TSTF)
Standard Technical Specification Change Traveler TSTF-65, Revision
1, approved by the NRC on November 10, 1994.
The proposed changes revise TS 5.0 to change management titles
from (a) ``Vice President, Diablo Canyon Operations and Plant
Manager'' to ``plant manager,'' (b) ``Senior Vice President and
General Manager--Nuclear Power Generation'' to ``specified corporate
officer,'' (c) ``Radiation Protection Director'' to ``radiation
protection manager,'' and (d) ``Operations Director'' to
``operations manager.''
The proposed changes do not eliminate any of the qualifications,
responsibilities or requirements for these positions. Each member of
the plant staff assigned to these positions shall continue to meet
or exceed the minimum qualifications of ANSI/ANS 3.1-1978,
Regulatory Guide 1.8, Revision 2, April 1987 (radiation protection
manager), or TS 5.2.2.e (operations manager) as required by TS
5.3.1.
The proposed change to replace the title ``Vice President,
Diablo Canyon Operations and Plant Manager'' with the generic title
``plant manager'' reflects PG&E's plan to split the responsibilities
of the Vice President, Diablo Canyon Operations and Plant Manager,
into two positions: (1) Vice President, Diablo Canyon Operations,
and (2) Station Director. The Station Director will report to the
Vice President, Diablo Canyon Operations. The Station Director will
fulfill the responsibilities of the ``Plant Manager'' as described
currently in TS and Final Safety Analysis Report (FSAR) Update and
will be responsible for overall safe operation of the plant and will
have control over those onsite activities necessary for safe
operation and maintenance of the plant. This change results in no
change to the responsibilities or qualification requirements for
this position as specified in the TS.
The remaining changes are administrative changes only that
result in no changes in the responsibilities for the positions.
None of the proposed changes have an impact on plant equipment,
or on how plant equipment is operated or maintained, and therefore
they have no impact on plant accidents.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes revise TS 5.0 to change management titles
from (a) ``Vice President, Diablo Canyon Operations and Plant
Manager'' to ``plant manager,'' (b) ``Senior Vice President and
General Manager--Nuclear Power Generation'' to ``specified corporate
officer,'' (c) ``Radiation Protection Director'' to ``radiation
protection manager,'' and (d) ``Operations Director'' to
``operations manager.''
The proposed changes do not eliminate any of the qualifications,
responsibilities or requirements for these positions.
None of the proposed changes have an impact on plant equipment,
or on how plant equipment is operated or maintained, and therefore
they have no impact on initiation of new or different plant
accidents.
Therefore, the proposed changes do 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 revise TS 5.0 to change management titles
from (a) ``Vice President, Diablo Canyon Operations and Plant
Manager'' to ``plant manager,'' (b) ``Senior Vice President and
General Manager--Nuclear Power Generation'' to ``specified corporate
officer,'' (c) ``Radiation Protection Director'' to ``radiation
protection manager,'' and (d) ``Operations Director'' to
``operations manager.''
The proposed changes do not eliminate any of the qualifications,
responsibilities or requirements for these positions.
None of the proposed changes have an impact on plant equipment,
or on how plant equipment is operated or maintained, and therefore
they have no impact on margin of safety.
Therefore, 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 requests involve no significant hazards consideration.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: December 7, 2000.
Description of amendment request: The licensee proposes to revise
Technical Specification 3.5.A.1 by adding a note regarding operability
of the Low Pressure Coolant Injection system (LPCI) under certain
restrictive conditions. The subject change would provide a
clarification of system operability that would result in additional
flexibility in operations during hot shutdown conditions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The LPCI system is not assumed to be the initiator of any
previously analyzed event. Its function is in mitigating and thereby
limiting consequences of analyzed events. With this proposed change
LPCI is still capable of being manually realigned, if needed, to
mitigate the consequences of accidents. The allowance provided by
this change is only applicable for the reactor in a shutdown
condition with reactor pressure less than the RHR [residual heat
removal] shutdown cooling permissive setpoint.
Thus, the reactor heat load is much less than assumed for design
basis loss of coolant accidents occurring at full power.
Furthermore, other emergency core cooling systems are still required
to be operable.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[[Page 7687]]
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change does not involve any physical alteration of
the plant or introduce new modes of operation. There is no change in
plant operation that involves failure modes other than those
previously evaluated.
The methods governing plant operation and testing remain
consistent with current safety analysis assumptions. Therefore, the
proposed change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
The proposed change has no impact on any safety analysis
assumption. The clarifying Note being added to Technical
Specification 3.5.A.1 allows the decay heat removal function to be
available without immediate shutdown requirements for inoperable
LPCI subsystems being imposed. This is recognition that the amount
of time to realign the RHR system from the decay heat removal
function has no significant impact on the margin of safety
associated with establishing LPCI injection, because the heat loads
under these conditions are far less than assumed in the safety
analysis.
Placing the reactor in SDC [Shutdown Cooling] during hot
shutdown is a normal and preferred method for removing sensible heat
from the reactor. In addition, the change does not alter the
availability of other safety systems and the ability to meet their
safety functions. The additional flexibility, to allow LPCI
subsystems to be considered operable during SDC below the RHR
shutdown cooling permissive pressure and without entering a shutdown
LCO [limiting condition for operation] will not significantly reduce
margins of safety since the reactor is in hot shutdown with all
control rods inserted, reactor pressure is less than the RHR
shutdown cooling permissive pressure, and other ECCS [Emergency Core
Cooling Systems] systems should be capable of providing the required
cooling, thereby allowing operation of RHR SDC when necessary. Thus,
the margins of safety for such situations are maintained.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: December 19, 2000.
Description of amendment request: The proposed change would revise
the reactor vessel pressure/temperature (P/T) limit curves specified in
TS 3.6.A.1, ``Reactor Coolant Systems--Pressure and Temperature
Limitations,'' as graphically represented in Figure 3.6.1, for reactor
heatup, cooldown, and critical operation, as well as for inservice
hydrostatic and leak tests.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The changes to the calculational methodology for the [pressure/
temperature] P/T limits based upon [American Society of Mechanical
Engineers] ASME [American Society for Mechanical Engineers Boiler
and Pressure Vessel Code] Code Cases N-640 and N-588 provide
adequate margin in the prevention of a brittle-type fracture of the
reactor pressure vessel (RPV). The Code Cases were developed based
upon the knowledge gained through years of industry experience. The
experience gained in the areas of fracture toughness of materials
and pre-existing undetected defects show that some of the existing
assumptions used for the calculation of P/T limits are unnecessarily
conservative and unrealistic. Therefore, providing the allowances of
the subject Code Cases in developing the P/T limit curves will
continue to provide adequate protection against nonductile-type
fractures of the RPV.
The evaluation for revising the P/T limit curves for
4.46 x 108MWH(t) (32 effective full power years) was
performed using the approved methodologies of 10 CFR 50, appendix G.
The curves generated from these methods ensure the P/T limits will
not be exceeded during any phase of reactor operation. The proposed
changes will not affect any other system or equipment designed for
the prevention or mitigation of previously analyzed events. Thus,
the probability of occurrence and the consequences of any previously
analyzed event are not significantly increased as the result of the
proposed changes.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed changes to the reactor pressure vessel P/T limits
do not affect the assumed performance of any system, structure, or
component previously evaluated. The proposed changes do not
introduce any new modes of system operation or failure mechanisms.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
Industry experience since the inception of the P/T limits in
1974 confirms that some of the existing methodologies used to
develop P/T curves is unnecessarily conservative. Accordingly, ASME
Code Cases N-640 and N-588 take advantage of the acquired knowledge
by establishing more enhanced methodologies for the development of
P/T curves. Therefore, operational flexibility can be gained without
a significant reduction in the margin of safety to RPV brittle
fracture.
The revised evaluation of the P/T curves to
4.46 x 108MWH(t) was performed per the guidelines of 10
CFR 50, and thus, the margin of safety is not reduced as the result
of the proposed changes.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: December 14, 2000
Description of amendment request: The proposed changes will modify
the Technical Specifications Section 3/4.7.7 ``Control Room Emergency
Habitability Systems'' Surveillance Requirements 4.7.7.1.d.1 and
4.7.7.2.a, to revise the differential pressure limit across the control
room emergency ventilation system filter assembly and increase the
minimum number of compressed air bottles in the control room bottled
air pressurization system.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[1.] Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Increasing the minimum required number of air bottles in the
control
[[Page 7688]]
room bottled air pressurization system in order to maintain system
capacity does not change the operation of the plant. The control room
bottled air pressurization system and the emergency ventilation system
will not be operated differently. No new accident initiators are
established as a result of the proposed changes. Revising the
differential pressure acceptance criteria and including [the] demister
filter along with the HEPA filter and charcoal adsorber will provide
increased assurance of system readiness. These systems will continue to
be operable to limit control room dose to within the analysis of
record. Therefore, the probabaility of occurrence or the consequences
of an accident previously evaluated is not increased.
[2.] Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes do not affect the operation of the plant. The
control room bottled air pressurization system and control room
emergency ventilation system will not be operated differently as a
result of the proposed changes. No new accident or event initiators are
being created by these changes. Therefore, the proposed changes do not
create the possibility of a new or different kind of accident from any
previously evaluated.
[3.] Involve a significant reduction in the margin of safety as
defined in the bases [of] any Technical Specifications.
The proposed changes reflect conservative changes in the operating
requirements for the control room bottled air pressurization and
control room emergency ventilation systems. These changes will further
ensure the systems will continue to be operable to mitigate the
consequence of an accident for the control room operators. Therefore,
the proposed changes do not result in 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Donald P. Irwin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Section Chief: Richard L. Emch, Jr.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852.
Publicly available records will be accessible electronically from the
ADAMS Public Library component on the NRC Web site, http://www.nrc.gov
(the Electronic Reading Room).
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of application for amendment: February 28, 2000, as
supplemented by letters dated May 12, May 24, June 1, and June 28,
2000.
Brief description of amendment: The amendment revised certain
license conditions to reflect the change in ownership interest from
PECO to Exelon Generation Company, LLC.
Date of issuance: January 12, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 228.
Facility Operating License No. DPR-50: Amendment revised the
License.
Date of initial notice in Federal Register: April 10, 2000 (65 FR
19029).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 21, 2000.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: October 12, 2000.
Brief description of amendment: The amendment changes the name of
the facility from WNP-2 to Columbia Generating Station in all
applicable locations of the Operating License, appendix A Technical
Specifications, and appendix B Environmental Protection Plan. In
addition, the proposed action would make editorial changes to TS Figure
4.1-1, ``Site Area Boundary'' modifying or deleting text associated
with references to WNP-2.
Date of issuance: January 8, 2001.
Effective date: January 8, 2001, and shall be implemented within 30
days from the date of issuance.
Amendment No: 169.
Facility Operating License No. NPF-21: The amendment revised the
Operating License, appendix A Technical Specifications, and appendix B
Environmental Protection Plan.
Date of initial notice in Federal Register: November 29, 2000 (65
FR 71134).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 8, 2001.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: November 23, 1999, as supplemented by
letters dated February 24 and October 19, 2000.
Brief description of amendment: The amendment incorporated the use
of American Society for Testing and Materials (ASTM) D3803-1989,
``Standard Test Method for Nuclear-Grade Activated Carbon,'' into the
Arkansas Nuclear One, Unit No. 1, Technical Specifications.
Date of issuance: December 28, 2000.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 210.
Facility Operating License No. DPR-51: Amendment revised the
Technical Specifications.
[[Page 7689]]
Date of initial notice in Federal Register: March 8, 2000 (65 FR
12291). The application was renoticed on March 22, 2000 (65 FR 15378).
The October 19, 2000, supplemental letter provided clarifying
information and revised Bases pages that was within the scope of the
application and did not change the associated no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 28, 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: June 10, 1999, as supplemented
by letters dated November 4, 1999, and October 12, 2000.
Brief description of amendment: By letter dated June 10, 1999,
FirstEnergy submitted its response for Davis-Besse Nuclear Power
Station to the actions requested in Generic Letter (GL) 99-02,
``Laboratory Testing of Nuclear-Grade Activated Charcoal,'' dated June
3, 1999. By letter dated November 4, 1999, FirstEnergy requested
changes to the Technical Specifications (TS) sections 3/4.6.4.4,
``Hydrogen Purge System (HPS),'' 3/4.6.5.1, ``Shield Building Emergency
Ventilation System (SBEVS),'' 3/4.7.6.1, ``Control Room Emergency
Ventilation System (CREVS),'' and 6.0, ``Administrative Controls,'' for
Davis-Besse Nuclear Power Station. FirstEnergy proposes adoption of a
Ventilation Filter Testing Program (VFTP) in TS section 6.0--
Administrative Control and removal of the specific ventilation filter
testing requirements from the plant's Surveillance Requirements of TS
sections 3/4.6.4.4, 3/4.6.5.1, and 3/4.7.6.1. By letter dated October
12, 2000, FirstEnergy provided additional information regarding
relative humidity in the control room. The proposed changes would
revise the TS surveillance testing of the safety related ventilation
system charcoal to meet the requested actions of GL 99-02.
Date of issuance: January 11, 2001.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 244.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 29, 1999 (64
FR 73091).
The supplemental information contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice. The Commission's related evaluation of the amendment
is contained in a Safety Evaluation dated January 11, 2001.
No significant hazards consideration comments received: No.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: June 28, 2000.
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.7.6.1, ``Plant Systems--Control Room Emergency
Ventilation System,'' to establish actions to be taken for an
inoperable control room ventilation system due to a degraded control
room boundary (CRB). This revision approves changes that would allow up
to 24 hours to restore the CRB to operable status when two control room
ventilation system trains are inoperable due to an inoperable CRB in
MODES 1, 2, 3, and 4. In addition, a Limiting Condition for Operation
note would be added to allow the CRB to be opened intermittently under
administrative controls without affecting control room ventilation
system operability. Various other editorial changes have been made to
reflect the revised TS. The applicable TS Bases have been revised to
document the TS changes and to provide supporting information.
Date of issuance: January 2, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 254.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 26, 2000 (65 FR
46010).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 2, 2001.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of application for amendment: November 10, 1999, as
supplemented October 3, 2000.
Brief description of amendment: The amendment revises Technical
Specification (TS) 5.5.7.c, to commit to the American Society for
Testing and Materials D3803-1989 test protocol for the ventilation
filter testing program. The changes are consistent with Nuclear
Regulatory Commission (NRC) Generic Letter 99-02.
Date of issuance: December 27, 2000.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 235.
Facility Operating License No. DPR-49: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 12, 2000 (65 FR
1924).
The supplemental information in the October 3, 2000, letter
contained clarifying information and did not change the initial no
significant hazards consideration determination and did not expand the
scope of the original Federal Register notice. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
December 27, 2000.
No significant hazards consideration comments received: No.
PECO Energy Company, PSEG Nuclear LLC, Delmarva Power and Light
Company, and Atlantic City Electric Company, Docket Nos. 50-277 and 50-
278, Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, York County,
Pennsylvania
Date of application for amendments: October 10, 2000.
Brief description of amendments: The amendments revised the
licenses for Peach Bottom Units 2 and 3 to remove Delmarva Power and
Light Company as a licensee, in conjunction with the transfer of the
minority ownership interests of Delmarva Power and Light Company to the
majority owners, PECO Energy Company and PSEG Nuclear LLC.
Date of issuance: December 29, 2000.
Effective date: As of date of issuance, to be implemented within 30
days.
Amendments Nos.: 238 & 241.
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the License.
Date of initial notice in Federal Register: November 27, 2000 (65
FR 70740).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 27, 2000.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of application for amendments: December 20, 1999, as
supplemented
[[Page 7690]]
February 11, February 25, and October 10, 2000.
Brief description of amendments: The amendments revised Facility
Operating Licenses DPR-70 and DPR-75 to reflect changes related to the
transfer of the license for the Salem Nuclear Generating Station, Unit
Nos. 1 and 2, to the extent held by Delmarva Power and Light Company,
to PSEG Nuclear Limited Liability Company.
Date of issuance: December 29, 2000.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment Nos.: 240 and 221.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the License.
Date of initial notice in Federal Register: February 18, 2000 (65
FR 8452). The February 11, February 25, and October 10, 2000,
supplements did not expand the scope of the original application with
respect to both the proposed transfer action and the proposed amendment
action as initially noticed in the Federal Register. No hearing
requests or comments were received. In addition, the submittal did not
affect the applicability of the Commission's generic no significant
hazards consideration determination set forth in 10 CFR 2.1315.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 21, 2000.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of application for amendments: October 13, 1999, as
supplemented by letter dated June 1, 2000.
Brief description of amendments: The amendments revised the
Technical Specifications to permit relaxation of allowed bypass test
times for Limiting Conditions for Operations (LCO) 3.3.1, ``Reactor
Trip System Instrumentation'', and LCO 3.3.2, ``Engineered Safety
Feature Actuation System Instrumentations''. These changes specifically
revise the completion times from 6 hours to 72 hours for inoperable
analog instruments, increase bypass times from 6 hours to 12 hours for
surveillance testing of analog channels, and increase completion times
from 6 hours to 24 hours for an inoperable logic cabinet or master and
slave relays.
Date of issuance: December 22, 2000.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 116 and 94.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 26, 2000 (65 FR
46016).
The supplemental letter dated June 1, 2000, provided clarifying
information that did not change the scope of the October 13, 1999,
application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 22, 2000.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 17th day of January 2001.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 01-1987 Filed 1-23-01; 8:45 am]
BILLING CODE 7590-01-P