, 5078-5088 [06-744]

Agencies

• Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations
• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 20 (Tuesday, January 31, 2006)]
[Notices]
[Pages 5078-5088]
From the Federal Register Online via the Government Printing 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, https://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, HearingDocket@nrc.gov; 
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 
OGCMailCenter@nrc.gov. 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, https://
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 pdr@nrc.gov.

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, https://
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 pdr@nrc.gov.

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 lic