[Federal Register Volume 71, Number 20 (Tuesday, January 31, 2006)]
[Notices]
[Pages 5078-5088]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-744]
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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 section 189a.(2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make
[[Page 5079]]
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 6, 2006 to January 19, 2006. The
last biweekly notice was published on January 17, 2006 (71 FR 2586).
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. Within 60 days after the date of publication of this
notice, 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.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place 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 Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, 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 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, 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 general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include 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/requestor to relief. A petitioner/
requestor who fails to satisfy 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.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no
[[Page 5080]]
significant hazards consideration, 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 by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
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 23, 2005.
Description of amendments request: The amendments would increase
the emergency diesel generator (EDG) allowed out of service time (AOT)
from 72 hours to 10 days, allow EDG starting air receiver pressure to
momentarily drop below limits during successful starting of an EDG, and
remove from the Technical Specifications the statement that the two
groups of pressurizer heaters are capable of being powered from an
emergency power supply.
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 consequences of an accident previously evaluated?
Response: No.
The proposed Technical Specification (TS) change to increase the
emergency diesel generator (EDG) allowed out of service time (AOT)
from 72 hours to 10 days will not cause an accident to occur and
will not result in any change in the operation of the associated
accident mitigation equipment. The EDGs are not accident initiators.
The EDGs are designed to mitigate the consequences of previously
evaluated accidents including a loss of offsite power. Extending the
AOT for a single EDG would not affect the previously evaluated
accidents since the remaining EDG supporting the redundant
Engineered Safety Features (ESF) systems would continue to be
available to perform the accident mitigation functions. The duration
of this TS AOT considers that there is a minimal possibility that an
accident will occur while a component is removed from service. A
risk informed assessment was performed which concluded that the
increase in plant risk is small and consistent with the guidance
contained in Regulatory Guide 1.177, ``An Approach for Plant-
Specific, Risk-Informed Decisionmaking: Technical Specifications.''
The design basis accidents will remain the same postulated events
described in the PVNGS [Palo Verde Nuclear Generating Station]
Updated Final Safety Analysis Report (UFSAR). In addition, extending
the EDG AOT will not impact the consequences of an accident
previously evaluated. The consequences of previously evaluated
accidents will remain the same during the proposed 10 day AOT as
during the current 72 hour AOT. The ability of the remaining TS-
required EDG to mitigate the consequences of an accident will not be
affected since no additional failures are postulated while equipment
is inoperable within the TS AOT. The remaining EDG is sufficient to
mitigate the consequences of any design basis accident.
The proposed addition of a note to Condition F of TS 3.8.3,
would allow EDG starting air receiver pressure to momentarily drop
below limits during successful starting of an EDG. The EDG air
starting system will not be operated or be configured any
differently than that which it is currently required and designed
for. This proposed change will only add a note for clarification to
Condition F of TS 3.8.3. This note describes entering this Condition
is not necessary when the EDG starts normally and is operating per
required procedures. Momentary transients outside the air receiver
pressure range do not invalidate the successful start and running of
the EDG. A successful start of the EDG indicates the starting air
system has performed its required safety function. This proposed
change will not increase the probability or consequence of an
accident previously evaluated.
The proposed TS change associated with the requirements for the
pressurizer heaters to be supplied by emergency power will not
result in any change in plant design. These components will continue
to be powered from Class 1E power sources as described in the
proposed TS Bases change associated with this change. As a result,
the operation and reliability of the pressurizer heaters will not be
affected by the proposed description change. In addition, operation
of the pressurizer heaters is not assumed to mitigate any design
basis accident. The proposed changes will not cause an accident to
occur and will not result in a change in the operation of any
accident mitigation equipment. The design basis accidents remain the
same postulated events described in the PVNGS UFSAR.
Therefore, the proposed changes do 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?
Response: No.
The proposed changes do not involve a change in the design,
configuration, or method of operation of the plant that could create
the possibility of a new or different [kind of] accident. Equipment
will be operated in the same configuration and manner that is
currently allowed and designed for. The proposed changes do not
introduce any new failure modes. This license amendment request does
not impact
[[Page 5081]]
any plant systems that are accident initiators or adversely impact
any accident mitigating systems.
Therefore, the proposed changes do 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?
Response: No.
The EDG reliability and availability are monitored and
evaluated, in accordance with 10 CFR 50.65 (Maintenance Rule)
performance criteria, to assure EDG out of service times do not
degrade operational safety over time. Extension of the EDG AOT will
not erode the reduction in severe accident risk that was achieved
with implementation of the Station Blackout (SBO) rule (10 CFR
50.63) or affect any safety analyses assumptions or inputs. The SBO
coping analysis is unaffected by the AOT extension since the EDGs
are not assumed to be available during the coping period. The
assumptions used in the coping analysis regarding EDG reliability
are unaffected since preventive maintenance and testing will
continue to be performed to maintain the reliability assumptions.
Accident mitigation functions will be maintained by the other
TS-required EDG availability to supply power to the safety related
Class 1E electrical loads. The availability of the TS-required
offsite power, combined with the availability of the PVNGS SBO Gas
Turbine Generators (GTGs) and the use of the Configuration Risk
Management Program required by 10 CFR 50.65(a)(4), provide adequate
compensation for the small incremental increase in plant risk of the
proposed EDG AOT extension. This small increase in plant risk while
operating is offset by a reduction in shutdown risk resulting from
the increased availability and reliability of the EDGs during
refueling outages, and avoiding transition risk incurred during
unplanned plant shutdowns. In addition, the calculated risk measures
associated with the proposed AOT are below the acceptance criteria
defined in Regulatory Guide 1.177.
The proposed change to add a note to Condition F of TS 3.8.3
does not involve changes to setpoints or limits established or
assumed by the accident analyses. This note only applies to those
occasions when after a successful start of an EDG has occurred and
the starting air receiver pressure has momentarily dropped below its
limit. This change allows for not declaring the EDG inoperable
solely due to this momentary drop in pressure during a successful
start of the EDG. No safety margin will be impacted by this change.
The proposed TS change associated with the wording description
of LCO [Limiting Condition of Operation] 3.4.9, ``Pressurizer,'' for
the requirement of the pressurizer heaters to be supplied by
emergency power does not adversely affect equipment design or
operation, and there are no changes being made to the TS-required
safety limits or system settings that would adversely affect plant
safety. The emergency power requirements for the pressurizer
heaters, which came from the Three Mile Island (TMI) action item
requirement II.E.3.1, ``Emergency Power Requirements for Pressurizer
Heater,'' of NUREG-0737, ``Clarification of TMI Action Plan
Requirements,'' will continue to be met. The pressurizer heaters
used to satisfy the NUREG-0737 and LCO 3.4.9 requirements are, by
design, permanently connected to Class 1E power supplies as
described in the PVNGS Updated Final Safety Analyses Report, Section
18.II.E.3.1.
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
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: Kenneth C. Manne, Senior Attorney, Arizona
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix,
Arizona 85072-2034.
NRC Branch Chief: David Terao.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: November 15, 2005.
Description of amendment request: The proposed change modifies the
technical specifications (TS) to clarify the wording of emergency
closed cooling water (ECCW) Surveillance Requirement (SR) 3.7.10.2. The
current wording in SR 3.7.10.2 requires that automatic valves on the
ECCW system actuate on an actuation signal. However, the TS Bases for
the SR identify more than just valves tested to include the automatic
start capability of the ECCW pump in each subsystem. Therefore, the
wording of this SR would be modified to clarify that its purpose is to
verify actuation of the entire subsystem on an actual or simulated
signal, rather than just verify valve actuation.
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.
There are no physical modifications being made to any plant
system or component. The only change is to a Surveillance
Requirement within the Technical Specifications, in order to improve
understanding and avoid misinterpretation of the requirements. The
original intent of ECCW SR 3.7.10.2 is maintained by the change
being proposed. The revised Technical Specification requirements do
not impact initiators of previously evaluated accidents or
transients.
The specification being revised is associated with a system used
to mitigate the consequences of accidents. The change to the wording
of ECCW SR 3.7.10.2 does not impact the capability of the associated
system to perform its required function. The reworded ECCW SR more
clearly requires that the system[']s total actuation capability be
maintained.
The change does not affect how plant systems are controlled or
operated or tested. The change continues to provide confirmation of
the capability of plant components to respond as required to
mitigate the consequences of events. Therefore, the proposed change
does 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.
There are no physical modifications being made to any plant
system or component, and the proposed change introduces no new
method of operation of the plant, or its systems or components.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The change to the ECCW SR continues to ensure the ECCW
subsystems are tested on the same periodicity to verify their
capability to respond to actuation signals from the Emergency Core
Cooling System (ECCS) Instrumentation Functions of Low Water Level
and High Drywell Pressure. Therefore, the necessary function of the
Technical Specification requirements is maintained, and 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: David W. Jenkins, Attorney, FirstEnergy
Corporation, Mail Stop A-GHE-107, 76 South Main Street, Akron, OH
44308.
NRC Branch Chief: Mindy Landau, Acting.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: December 13, 2005.
Description of amendment request: The proposed amendments would
revise technical specification (TS) requirements for surveillance
[[Page 5082]]
requirements for containment integrated leakage rate testing in TS
5.5.14.a to allow a one-time extension of the interval between reactor
containment vessel integrated leakage rate tests (ILRTs) from 10 to 15
years.
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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
This license amendment proposes to revise the Technical
Specifications to allow for the one time extension of the
containment integrated leakage rate test interval from 10 to 15
years. The containment vessel function is purely mitigative. There
are no design basis accidents initiated by a failure of the
containment leakage mitigation function. The extension of the
containment integrated leakage rate test interval will not create
any adverse interactions with other systems that could result in
initiation of a design basis accident. Therefore, the probability of
occurrence of an accident previously evaluated is not significantly
increased.
The potential consequences of the proposed change have been
quantified by analyzing the changes in risk that would result from
extending the containment integrated leakage rate test interval from
10 to 15 years. The increase in risk in terms of person-rem per year
within 50 miles resulting from design basis accidents was estimated
to be of a magnitude that NUREG-1493, ``Performance-Based
Containment Leak-Test Program'', indicates is imperceptible. The
Nuclear Management Company has also analyzed the increase in risk in
terms of the frequency of large early releases from accidents. The
increase in the large early release frequency resulting from the
proposed extension was determined to be within the guidelines
published in Regulatory Guide 1.174, ``An Approach for Using
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Current Licensing Basis''. Additionally, the
proposed change maintains defense-in-depth by preserving a
reasonable balance among prevention of core damage, prevention of
containment failure, and consequence mitigation. The Nuclear
Management Company has determined that the increase in conditional
containment failure probability from reducing the containment
integrated leakage rate test frequency from 1 test per 10 years to 1
test per 15 years would be small.
Continued containment integrity is also assured by the history
of successful containment integrated leakage rate tests, and the
established programs for local leakage rate testing and in-service
inspections which are unaffected by the proposed change. Therefore,
the probability of occurrence or the consequences of an accident
previously analyzed are not significantly increased.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change to extend the containment integrated leakage
rate test interval from 10 to 15 years does not create any new or
different accident initiators or precursors. The length of the
containment integrated leakage rate test interval does not affect
the manner in which any accident begins. The proposed change does
not create any new failure modes for the containment and does not
affect the interaction between the containment and any other system.
Thus, the proposed changes do not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The risk-based margins of safety associated with the containment
integrated leakage rate test are those associated with the estimated
person-rem per year, the large early release frequency, and the
conditional containment failure probability. The Nuclear Management
Company has quantified the potential effect of the proposed change
on these parameters and determined that the effect is not
significant. The non-risk-based margins of safety associated with
the containment integrated leakage rate test are those involved with
its structural integrity and leak tightness. The proposed change to
extend the containment integrated leakage rate test interval from 10
to 15 years does not adversely affect either of these attributes.
The proposed change only affects the frequency at which these
attributes are verified. Therefore, the proposed change does not
involve a significant reduction in margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Acting Branch Chief: Timothy Kobetz.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: November 18, 2005.
Description of amendment request: The proposed amendment would
change the SSES 1 and 2 Technical Specifications (TSs) to implement the
Average Power Range Monitor/Rod Block Monitor/Technical Specifications/
Maximum Extended Load Line Limit Analysis (ARTS/MELLLA). Specifically,
the average power range monitor (APRM) flow-biased scram and rod block
trip setpoints would be revised to permit operation in the MELLLA
region. The current flow-biased rod block monitor (RBM) would also be
replaced by a power dependent RBM implemented through the referenced
proposed upgrade to a digital power range neutron monitor system
(PRNMS). The change from the flow-biased RBM to the power-dependent RBM
would also require new trip setpoints. In addition, the flow-biased
APRM scram and rod block trip setdown requirement would be replaced by
more direct power and flow-dependent thermal limits to reduce the need
for APRM gain adjustments, and to allow more direct thermal limits
administration during operation other than rated conditions. Finally,
the proposed amendment would change the methods used to evaluate the
annulus pressurization (AP), mass blowdown, and early release resulting
from the postulated recirculation suction line break (RSLB).
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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
Proposed Change No. 1: The proposed change eliminates the
Average Power Range Monitor (APRM) flow-biased scram and rod block
trip setpoint setdown requirements and substitutes power and flow
dependent adjustments to the Minimum Critical Power Ratio (MCPR) and
Linear Heat Generation Rate (LHGR) thermal limits. Thermal limits
will be determined using NRC approved analytical methods. The
proposed change will have no effect upon any accident initiating
mechanism. The power and flow dependent adjustments will ensure that
the MCPR safety limit will not be violated as a result of any
Anticipated Operational Occurrence (AOO), and that the fuel thermal
and mechanical design bases will be maintained. Therefore, the
proposed change will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Proposed Change No. 2: The proposed change expands the power and
flow operating domain by relaxing the restrictions imposed by the
formulation of the APRM flow-biased scram and rod block trip
setpoints and the replacement of the current flow-biased RBM with a
new power dependent RBM, which will be implemented using a digital
Power Range Neutron Monitoring System (PRNMS). The APRM and RBM are
not involved in the initiation of any
[[Page 5083]]
accident; and the APRM flow-biased scram and rod block functions are
not credited in any PPL safety licensing analyses.
The analysis of the instrument line break event resulted in an
insignificant change in the radiological consequences. The change
for the instrument line break was an insignificant increase of 0.1
Rem.
Since the proposed changes will not affect any accident
initiator, or introduce and initial conditions that would result in
NRC approved criteria being exceeded, and since the APRM and RBM
will remain capable of performing their design functions, the
proposed change will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Proposed Change No. 3: The methods used to evaluate Annulus
Pressurization (AP) and mass blowdown and energy releases resulting
from the postulated Recirculation Suction Line Break (RSLB) at the
MELLLA conditions are changed to use more realistic, but still
conservative, methods of analysis to determine an AP mass and energy
release profile for AP loads resulting from the postulated RSLB. The
releases resulting from the RSLB at off-rated conditions have been
demonstrated to be bounded by the current design basis loads. Since
the proposed changes do not affect any accident initiator and since
the RSLB AP releases remain bounded by the current design basis, the
proposed changes do not involve a significant increase in the
probability or radiological consequences of an accident previously
evaluated. Therefore the proposed changes do not involve a
significant increase in the probability or consequences of any
accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Proposed Change No. 1: The proposed change eliminates the
Average Power Range Monitor (APRM) flow-biased scram and rod block
setpoint setdown requirements and substitutes power and flow
dependent adjustments to the Minimum Critical Power Ratio (MCPR) and
Linear Heat Generation Rate (LHGR) thermal limits. Because the
thermal limits will continue to be met, no analyzed transient event
will escalate into a new or different type of accident due to the
initial starting conditions permitted by the adjusted thermal
limits. Therefore, the proposed change will not create the
possibility of a new or different kind of accident previously
evaluated.
Proposed Change No. 2: The proposed change expands the power and
flow operating domain by relaxing the restrictions imposed by the
formulation of the APRM flow-biased scram and rod block trip
setpoints and the replacement of the current flow-biased RBM with a
new power dependent RBM, which will be implemented using a digital
Power Range Neutron Monitoring System (PRNMS). Changing the
formulation for the APRM flow-biased scram and rod block trip
setpoints and from a flow-biased RBM to a power dependent RBM does
not change their respective functions and manner of operation. The
change does not introduce a sequence of events or introduce a new
failure mode that would create a new or different type of accident.
The APRM flow-biased rod block trip setpoint will continue to block
control rod withdrawal when core power significantly exceeds normal
limits and approaches the scram level. The APRM flow-biased scram
trip setpoint will continue to initiate a scram if the increasing
power/flow condition continue beyond the APRM flow-biased rod block
setpoint. The power dependent RBM will prevent rod withdrawal when
the power dependent RBM rod block setpoint is reached. No new
failure mechanisms, malfunctions, or accident initiators are being
introduced by the proposed changes. In addition, operating within
the expanded power flow map will not require any systems, structures
or components to function differently than previously evaluated and
will not create initial conditions that would result in a new or
different kind of accident from any accident previously evaluated.
Proposed Change No. 3: The methods used to evaluate Annulus
Pressurization (AP) and mass blowdown and energy releases resulting
from the postulated Recirculation Suction Line Break (RSLB) at the
MELLLA conditions are changed to use more realistic, but still
conservative, methods of analysis to determine an AP mass and energy
release profile for AP loads resulting from the postulated RSLB. The
proposed changes to the methods of analysis to determine AP mass and
energy releases resulting from the postulated RSLB do not change the
design function or operation of any plant equipment. No new failure
mechanisms, malfunctions, or accident initiators are being
introduced by the proposed changes. Therefore, the proposed changes
do not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
Proposed Change No. 1: The proposed change eliminates the
Average Power Range Monitor (APRM) flow-biased scram and rod block
setpoint setdown requirements and substitutes power and flow
dependent adjustments to the Minimum Critical Power Ratio (MCPR) and
Linear Heat Generation Rate (LHGR) thermal limits. Replacement of
the APRM setpoint setdown requirement with power and flow dependent
adjustments to the MPR and LHGR thermal limits will ensure that
margins to the fuel cladding Safety Limit are preserved during
operation at other than rated conditions. Thermal limits will be
determined using NRC approved analytical methods. The power and flow
dependent adjustments will ensure that the MPR safety limit will not
be violated as a result of any Anticipated Operational Occurrence
(AOO), and that the fuel thermal and mechanical design bases will be
maintained. The 10 CFR 50.46 acceptance criteria for the performance
of the Emergency Core Cooling System (ECCS) following postulated
Loss-Of-Coolant Accidents (LOCAs) will continue to be met.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
Proposed Change No. 2: The proposed change expands the power and
flow operating domain by relaxing the restrictions imposed by the
formulation of the APRM flow-biased scram and rod block trip
setpoints and the replacement of the current flow-biased RBM with a
new power dependent RBM, which will be implemented using a digital
Power Range Neutron Monitoring System (PRNMS). The APRM flow-biased
rod block trip setpoint will continue to block control rod
withdrawal when core power significantly exceeds normal limits and
approaches the scram level. The APRM flow-biased scram trip setpoint
will continue to initiate a scram if the increasing power/flow
condition continues beyond the APRM flow-biased rod block setpoint.
The RBM will continue to prevent rod withdrawal when the power
dependent RBM rod block setpoint is reached. The MPR and LHGR
thermal limits will be developed to ensure that fuel thermal
mechanical design bases shall remain within the licensing limits
during a rod withdrawal error event and to ensure that the MPR
safety limit will not be violated as a result of a rod withdrawal
error event. Operation in the expanded operating domain will not
alter the manner in which safety limits, limiting safety system
settings, or limiting conditions for operation are determined.
Anticipated operational occurrences and postulated accident within
the expanded operating domain will be evaluated using NRC approved
methods. Therefore, the proposed change will not involve a
significant reduction in the margin of safety.
Proposed Change No. 3: The methods used to evaluate Annulus
Pressurization (AP) and mass blowdown and energy releases resulting
from the postulated Recirculation Suction Line Break (RSLB) at the
MELLLA conditions are changed to use more realistic, but still
conservative, methods of analysis to determine an AP mass and energy
release profile for AP loads resulting from the postulated RSLB.
Mass and energy releases for AP loads resulting from the postulated
RSLB remain bounded by the current design basis releases. 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: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief: Richard J. Laufer.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: October 11, 2005.
[[Page 5084]]
Description of amendment request: The proposed amendment would
remove the Technical Specification (TS) 3.1.5 requirement for the
Standby Liquid Control (SLC) system to be operable in Operational
Condition 5 (refueling) with any control rod withdrawn. Corresponding
changes would also be made to the SLC Initiation sections of Tables
3.3.2-1 and 4.3.2-1.
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:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to delete the operability requirement for
the SLC System in OPERATIONAL CONDITION 5* (OPERATIONAL CONDITION 5
with any control rod withdrawn) does not affect the probability or
consequences of an accident previously evaluated. In STARTUP and
POWER OPERATION, the SLC System is required to provide shutdown
capability. In HOT SHUTDOWN and COLD SHUTDOWN, control rods are not
able to be withdrawn since the reactor mode switch is in Shutdown
and a control rod block is applied. This provides adequate controls
to ensure that the reactor remains subcritical. Design basis
accident mitigation scenarios for OPERATIONAL CONDITION 5 do not
depend on, or require, SLC System operability. In REFUELING mode,
only a single control rod can be withdrawn from a core cell
containing fuel assemblies. Demonstration of adequate shutdown
margin in accordance with TS LIMITING CONDITION FOR OPERATION 3.1.1
ensures that the reactor will not become critical. Since the purpose
of the SLC System is to bring the reactor to a cold shutdown
condition from normal power operations and maintain it in a cold
shutdown condition, there is no design basis for the SLC System to
be required to be OPERABLE when only a single control rod can be
withdrawn. In addition, the reactor protection system and the
control rod system would continue to be able to provide protection
in the unlikely event that an inadvertent criticality occurs.
Therefore, these changes do 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?
Response: No.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated in
the UFSAR [updated final safety analysis report]. No new accident
scenarios, failure mechanisms, or limiting single failures are
introduced as a result of the proposed changes. Specifically, no new
hardware is being added to the plant as part of the proposed change,
no existing equipment is being modified, and no significant changes
in operations are being introduced.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes will not alter any assumptions, initial
conditions, or results of any accident analyses. The purpose of the
SLC System is to bring the reactor to and maintain it in a cold
shutdown condition following a failure to scram during plant
operations. The SLC System is not designed to terminate an
inadvertent criticality during REFUELING. Shutdown margin, either
demonstrated or analytically determined, in accordance with
Technical Specifications and procedural controls, will assure that
an inadvertent criticality event will not occur during REFUELING. In
addition, the reactor protection system and control rod system
provide protection in the unlikely event that an inadvertent
criticality occurs. The proposed change does not affect the ability
of the SLC System to achieve plant shutdown under analyzed
conditions (POWER OPERATION and STARTUP).
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Darrell J. Roberts.
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 (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of application for amendments: October 5, 2004, as
supplemented March 22, August 29, and October 31, 2005.
Brief description of amendments: The amendments revised the BVPS-1
and 2 Technical Specifications (TSs) 3/4.3.1, ``Reactor Trip System
Instrumentation,'' and 3/4.3.2, ``Engineered Safety Feature Actuation
Instrumentation,'' to modify steam generator (SG) level allowable value
(AV) setpoints. Specifically, the TS changes increased the AVs of the
SG water level-low-low setpoints from 14.6 percent and 16 percent to
19.6 percent and 20 percent of the narrow range (NR) instrument span
for BVPS-1 and 2, respectively. These are the AVs of setpoints
specified in TS Table 3.3-1 to initiate a reactor trip, and the
actuation setpoints specified in TS Table 3.3-3 to
[[Page 5085]]
start the auxiliary feedwater pumps. Also, for BVPS-2, the AV of the SG
water level-high-high setpoint increased from 81.1 percent to 92.7
percent of the NR span. This is the AV of a setpoint for actuation of
the turbine trip and the feedwater system isolation specified in TS
Table 3.3-3.
Date of issuance: January 11, 2006.
Effective date: Upon issuance and shall be implemented within 60
days.
Amendment Nos.: 270 and 152.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 23, 2004 (69
FR 68183). The supplements dated March 22, August 29, and October 31,
2005, provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the Nuclear Regulatory Commission staff's original
proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 11, 2006.
No significant hazards consideration comments received: No.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: January 10, 2005.
Description of amendment request: The amendment revised the
Seabrook Station, Unit No. 1, Technical Specifications (TSs) to extend
the interval for the performance of Containment Air Lock Interlock
Surveillance Requirement 4.6.1.3 from 6 months to 24 months.
Date of issuance: January 6, 2006.
Effective date: As of its date of issuance, and shall be
implemented within 30 days.
Amendment No.: 106.
Facility Operating License No. NPF-86: The amendment revised the
TSs.
Date of initial notice in Federal Register: May 24, 2005 (70 FR
29796). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 6, 2006.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: September 1, 2005.
Brief description of amendments: The amendments delete the
Technical Specification requirements for Occupational Radiation
Exposure Reports and Monthly Operating Reports.
Date of Issuance: January 13, 2006.
Effective Date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 198 and 141.
Renewed Facility Operating License Nos. DPR-67 and NPF-16:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 25, 2005 (70 FR
61661). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 13, 2006.
No significant hazards consideration comments received: No.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of application for amendments: July 21, 2005.
Brief description of amendments: The amendments delete the
Technical Specification requirements for Occupational Radiation
Exposure Reports and Monthly Operating Reports.
Date of issuance: January 13, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos: 228 and 224.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 25, 2005 (70 FR
61660).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 13, 2006.
No significant hazards consideration comments received: No.
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 application for amendments: July 29, 2005.
Brief description of amendments: The amendments revise the units'
Technical Specifications by eliminating the requirements to submit
monthly operating reports and occupational radiation exposure reports.
Date of issuance: January 12, 2006.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 292, 274.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 6, 2005 (70 FR
72673). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 12, 2006.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: March 8, 2005, as supplemented by letter
dated August 18, 2005.
Brief description of amendment: The amendment revised the Technical
Specification 2.1.1.2 for the single recirculation loop Safety Limit
Minimum Critical Power Ratio value to reflect results of a cycle-
specific calculation.
Date of issuance: January 4, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 215.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 29, 2005 (70 FR
15944). The supplement dated August 18, 2005, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 4, 2006.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: July 21, 2005.
Brief description of amendment: The amendment revises the technical
specifications testing frequency for the surveillance requirement (SR)
in TS 3.1.4, ``Control Rod Scram Times.'' Specifically, the proposed
change would revise the frequency for SR 3.1.4.2, control rod scram
time testing, from ``120 days cumulative operation in MODE 1'' to ``200
days cumulative operation in MODE 1.''
Date of issuance: January 5, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 216.
[[Page 5086]]
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 25, 2005 (70 FR
61661). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 5, 2006.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: June 30, 2004.
Brief description of amendment: The amendment revised Table 4.2.1,
``Minimum Test and Calibration Frequency for Core Cooling, Rod Block
and Isolation Instrumentation,'' of the Technical Specifications to
shorten the test interval between surveillance tests for the scram
discharge volume high level rod block, and the safety/relief valve low-
low set logic inhibit timer.
Date of issuance: January 12, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 144.
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 18, 2005 (70 FR
2892). The supplemental letters 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 12, 2006.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: January 11, 2005.
Brief description of amendment: The amendment deletes requirements
from the Technical Specifications for annual Occupational Radiation
Exposure Reports and Monthly Operating Reports.
Date of issuance: January 11, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 161.
Facility Operating License No. NPF-57: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 29, 2005 (70 FR
15946). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 11, 2006.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: February 25, 2005.
Brief description of amendment: The amendment revised Technical
Specification 3.1.3.1, ``Control Rod Operability,'' for the condition
of having one or more scram discharge volume vents or drain lines with
inoperable valves.
Date of issuance: January 13, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 162.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 7, 2005 (70 FR
33217). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 13, 2006.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, 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: January 11, 2005.
Brief description of amendments: The amendments deleted
requirements from the Technical Specifications (TSs) for annual
Occupational Radiation Exposure Reports and Monthly Operating Reports.
Date of issuance: January 11, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos.: 270 and 251.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the TSs.
Date of initial notice in Federal Register: March 29, 2005 (70 FR
15946) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 11, 2006.
No significant hazards consideration comments received: No
PSEG Nuclear, LLC, 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: February 15, 2005.
Brief description of amendments: These amendments delete the total
water and steam volume of the reactor coolant system from TS 5.4.2.
Date of issuance: January 11, 2006.
Effective date: As of the date of issuance and to be implemented
within 60 days.
Amendment Nos.: 269 and 250.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the TSs.
Date of initial notice in Federal Register: March 29, 2005 (70 FR
15940). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 11, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: April 4, 2005, as supplemented
by letters dated September 30 and November 8, 2005.
Brief description of amendment: The amendment supports the steam
generator replacement project by temporarily allowing one of the shield
building dome penetrations to be opened up to five hours a day, six
days a week while in Modes 1-4 during Cycle 7 operation until entering
Mode 5 at the start of the Cycle 7 refueling outage in fall 2006.
Date of issuance: January 6, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 59.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 19, 2005 (70 FR
41446). The supplemental letters provided clarifying information that
was within the scope of the initial notice and did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 6, 2006.
No significant hazards consideration comments received: No.
[[Page 5087]]
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room (PDR), located at One White Flint North, Public File Area 01F21,
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly
available records will be accessible from the Agencywide Documents
Access and Management System's (ADAMS) Public Electronic Reading Room
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the PDR Reference
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, 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 2.309, which
is available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a
request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, 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 general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
[[Page 5088]]
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. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ 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/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
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\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
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Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
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. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer or the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
1 (ANO-1), Pope County, Arkansas
Date of amendment request: January 3, 2006, as supplemented by
letters dated January 6 and 10, 2006.
Description of amendment request: Entergy Operations, Inc.
(Entergy) requests an emergency Technical Specification (TS) change to
the Steam Generator Level--Low allowable value of Limiting Condition
for Operation 3.3.11, ``Emergency Feedwater [EFW] Initiation and
Control (EFIC) System Instrumentation.'' Operation at 100 percent power
with the current allowable value involves an increased risk of spurious
EFW initiation. Therefore, Entergy requests a revised TS allowable
value of >= 9.34 inches and a limiting trip setpoint value of >= 10.42
inches in order to achieve and maintain 100 percent power operation. An
actuation time delay of <= 10.4 seconds is also proposed to minimize
the possibility of inadvertent actuations during anticipated transients
such as main feedwater transients or main turbine trips.
Date of issuance: January 13, 2006.
Effective date: As of the date of issuance and shall be implemented
within 7 days from the date of issuance.
Amendment No.: 227.
Renewed Facility Operating License No. DPR-51: Amendment revised
the Technical Specification.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No. The Commission's related evaluation of the
amendment, finding of emergency circumstances, state consultation, and
final NSHC determination are contained in a safety evaluation dated
January 13, 2006.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Stawn, 1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
Dated at Rockville, Maryland, this 20th day of January 2006.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 06-744 Filed 1-30-06; 8:45 am]
BILLING CODE 7590-01-P