[Federal Register Volume 71, Number 118 (Tuesday, June 20, 2006)]
[Notices]
[Pages 35456-35466]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-9434]
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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 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 May 25, 2006 to June 8, 2006. The last
biweekly notice was published on June 6, 2006 (71 FR 32603).
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
[[Page 35457]]
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
O1F21, 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 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,
[[Page 35458]]
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].
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: May 8, 2006.
Description of amendment request: The proposed change will add an
NRC-approved topical report to the analytical methods referenced in
Technical Specification (TS) Section 5.6.5, ``Core Operating Limits
Report (COLR).''
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.
Core operating limits are established each operating cycle in
accordance with TS 3.2, ``Power Distribution'' and TS 5.6.5, ``Core
Operating Limits Report (COLR)''. These core operating limits ensure
that the fuel design limits are not exceeded during any conditions
of normal operation or in the event of any Anticipated Operational
Occurrence (AOO). In addition, the Average Planar Linear Heat
Generation Rate (APLHGR) operating limits imposed by Technical
Specification 3.2.1 also ensure that the Peak Cladding Temperature
(PCT) during the postulated design[-]basis LOCA [loss-of-coolant
accident] does not exceed the 2200 [deg]F limit specified in 10 CFR
50.46. The APLHGR is a measure of the average linear heat generation
rate of all the fuel rods in a fuel assembly at any axial location.
The methods used to determine the operating limits are those
previously found acceptable by the NRC and listed in TS Section
5.6.5.b. A change to TS Section 5.6.5.b is requested to include an
updated LOCA analysis method, EXEM BWR-2000. The updated method will
be used to determine the APLHGR operating limits imposed by
Technical Specification 3.2.1. EXEM BWR-2000 has been reviewed and
approved by the NRC and is applicable to the GGNS [Grand Gulf
Nuclear Station, Unit 1] plant design and the FRA-ANP [Framatome-
Advance Nuclear Power] fuel being used at GGNS. The application of
the LOCA analytical model will continue to ensure that the APLHGR
operating limits are established to protect the fuel cladding
integrity during normal operation, AOOs, and the design-basis LOCA.
The requested TS changes concern the use of analytical methods and
do not involve any plant modifications or operational changes that
could affect any postulated accident precursors or accident
mitigation systems and do not introduce any new accident initiation
mechanisms.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed TS amendment will not change the design function,
reliability, performance, or operation of any plant systems,
components, or structures. It does not create the possibility of a
new failure mechanism, malfunction, or accident initiators not
considered in the design and licensing bases. Plant operation will
continue to be within the core operating limits that are established
using NRC[-]approved methods that are applicable to the GGNS design
and the GGNS fuel.
Therefore, the proposed change does 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 ECCS [emergency core cooling system] performance analysis
methods are used to establish the APLHGR limits required by
Technical Specification 3.2.1. The APLHGR limits are specified in
the COLR and are the result of fuel design, design[-]basis accident
(DBA), and transient analyses. Limits on the APLHGR are specified to
ensure that the fuel design limits are not exceeded during
anticipated operational occurrences (AOOs) and that the peak
cladding temperature (PCT) during the postulated design[-]basis LOCA
does not exceed the 2200 [deg]F limit specified in 10 CFR 50.46.
The EXEM BWR-2000 evaluation model is an updated LOCA analytical
method that has been approved by the NRC and is applicable to the
GGNS plant design and the fuel being used at GGNS. A GGNS plant[-
]specific ECCS performance analysis has been performed with the EXEM
BWR-2000 evaluation model. This evaluation concluded that the
resulting PCT still afforded adequate margin to the 2200 [deg]F
limit of 10 CFR 50.46.
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: Nicholas S. Reynolds, Esquire, Winston and
Strawn LLP, 1700 K Street, NW., Washington, DC 20006
NRC Branch Chief: David Terao.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of amendment request: May 24, 2006.
Description of amendment request: This amendment revises TS 1.0,
Definitions, TS 3/4.4.5, Steam Generator Tube Integrity, TS 3/4.4.6.2,
Reactor Coolant System (RCS) Operational LEAKAGE, adds a new
specification TS 6.8.4.k for Steam Generator Program and adds a new TS
6.9.1.12, Steam Generator Tube Inspection Report. The proposed changes
are necessary in order to implement the guidance for the industry
initiative on NEI 97-06, ``Steam Generator Program Guidelines.'' The
NRC staff issued a notice of availability of a model safety evaluation
and model no significant hazards consideration (NSHC) determination for
referencing in license amendment applications in the Federal Register
on March 2, 2005, (70 FR 10298). The licensee affirmed the
applicability of the model NSHC determination in its application dated
May 24, 2006.
[[Page 35459]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change requires a SG Program that includes
performance criteria that will provide reasonable assurance that the
SG tubing will retain integrity over the full range of operating
conditions (including startup, operation in the power range, hot
standby, cooldown and all anticipated transients included in the
design specification). The SG performance criteria are based on tube
structural integrity, accident induced leakage, and operational
LEAKAGE.
A SGTR event is one of the design basis accidents that are
analyzed as part of a plant's licensing basis. In the analysis of a
SGTR event, a bounding primary to secondary LEAKAGE rate equal to
the operational LEAKAGE rate limits in the licensing basis plus the
LEAKAGE rate associated with a double-ended rupture of a single tube
is assumed.
For other design basis accidents such as MSLB, rod ejection, and
reactor coolant pump locked rotor the tubes are assumed to retain
their structural integrity (i.e., they are assumed not to rupture).
These analyses typically assume that primary to secondary LEAKAGE
for all SGs is 1 gallon per minute or increases to 1 gallon per
minute as a result of accident induced stresses. The accident
induced leakage criterion introduced by the proposed changes
accounts for tubes that may leak during design basis accidents. The
accident induced leakage criterion limits this leakage to no more
than the value assumed in the accident analysis.
The SG performance criteria proposed change to the TS identify
the standards against which tube integrity is to be measured.
Meeting the performance criteria provides reasonable assurance that
the SG tubing will remain capable of fulfilling its specific safety
function of maintaining reactor coolant pressure boundary integrity
throughout each operating cycle and in the unlikely event of a
design basis accident. The performance criteria are only a part of
the SG Program required by the proposed change to the TS. The
program, defined by NEI 97-06, Steam Generator Program Guidelines,
includes a framework that incorporates a balance of prevention,
inspection, evaluation, repair, and leakage monitoring. The proposed
changes do not, therefore, significantly increase the probability of
an accident previously evaluated.
The consequences of design basis accidents are, in part,
functions of the DOSE EQUIVALENT 1-131 in the primary coolant and
the primary to secondary LEAKAGE rates resulting from an accident.
Therefore, limits are included in the plant technical specifications
for operational leakage and for DOSE EQUIVALENT 1-131 in primary
coolant to ensure the plant is operated within its analyzed
condition. The typical analysis of the limiting design basis
accident assumes that primary to secondary leak rate after the
accident is 1 gallon per minute with no more than 150 gallons per
day in any one SG, and that the reactor coolant activity levels of
DOSE EQUIVALENT 1-131 are at the TS values before the accident. The
proposed change does not affect the design of the SGs, their method
of operation, or primary coolant chemistry controls. The proposed
approach updates the current TSs and enhances the requirements for
SG inspections. The proposed change does not adversely impact any
other previously evaluated design basis accident and is an
improvement over the current TSs.
Therefore, the proposed change does not affect the consequences
of a SGTR accident and the probability of such an accident is
reduced. In addition, the proposed changes do not affect the
consequences of an MSLB, rod ejection, or a reactor coolant pump
locked rotor event, or other previously evaluated accident.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed performance based requirements are an improvement
over the requirements imposed by the current technical
specifications. Implementation of the proposed SG Program will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
result of the implementation of the SG Program will be an
enhancement of SG tube performance. Primary to secondary LEAKAGE
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the SG Program to manage SG tube
inspection, assessment, repair, and plugging. The requirements
established by the SG Program are consistent with those in the
applicable design codes and standards and are an improvement over
the requirements in the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TS.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. Hagood Hamilton, South Carolina Electric
& Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Section Chief: Evangelos C. Marinos.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: May 1, 2006 (TS-05-10).
Description of amendment request: The proposed amendment would
extend the burnup limit of the Mark-BW fuel design with advanced alloy
material referred to as M5 alloy. This proposed change affects Section
6.9.1.14.a of the Sequoyah Nuclear Plant Technical Specifications
(TSs). The impact to Section 6.9.1.14.a includes adding an NRC-approved
topical report (TR) associated with M5 alloy fuel assemblies. This TR
will be utilized, among others, in the determination of core operating
limits for each fuel cycle. In addition, the proposed amendment
includes the adoption of Industry/Technical Specification Task Force
(TSTF) Traveler, TSTF-363, Revision 0, ``Revised Topical Report
References in Improved Technical Specification (ITS) 5.6.5, Core
Operating Limits Report (COLR),'' which removes any references to
dates, revision numbers, and supplements in the TS listing of TRs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the
[[Page 35460]]
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.
In general, fuel assemblies and more specifically fuel rod
cladding, of any burnup level, is not a precursor to accidents
previously evaluated. An evaluation has been performed of the Mark-
BW design fuel assembly for all loss-of-coolant accidents (LOCA) and
non-LOCA transient events. This evaluation confirmed and justified
the use of Mark-BW fuel for operation in Sequoyah Nuclear Plant
(SQN) Units 1 and 2.
The ability of the M5 fuel rod cladding material to provide a
barrier against the release of radioactive fuel material has not
been reduced with respect to the Zircaloy-4 material. The approved
TR evaluated postulated accidents that involved adverse core
conditions and the release of radionuclides, and found that higher
burnup limits have very little impact on the overall radiological
consequences. Radiological consequences, as well as other safety
limits, are evaluated on a cycle-to-cycle basis to confirm that the
analyses of record remain bounding. If a proposed extended burnup
core design exceeds bounding safety analysis values, then either the
core design would be changed, or the safety values would be changed.
Rod cladding failures are assumed to occur in the fuel handling
accident; however, the consequences of this event are independent of
the properties of the fuel rod cladding. This is based on the fuel
handling event assuming the rupture of all fuel rods regardless of
the rod cladding material.
No change is proposed to the established safety analysis fuel
assembly inputs, specifically fuel assemblies are still limited to a
maximum 1500 effective full power day (EFPD) burnup and the reactor
core average maximum burnup will remain at 1000 EFPD burnup ensuring
the present accident analyses remain bounding. Based on above
discussion, the proposed revision to extend the burnup limit of M5
fuel rod cladding material will not significantly increase the
consequences of an accident and the potential for the release of
radioactive material to the environment.
Removing revision numbers, dates, and parenthetical information
from the listed TRs has no impact on the actual analytical methods
used to determine the core operating limits, nor does the change
have impact on the calculations performed for the current or future
reloads. This change is administrative in nature. This change has no
impact on plant equipment operation nor does it affect the
likelihood or consequences of an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Mark-BW fuel design with M5 alloy has been demonstrated to have
similar characteristics to that of the Mark-B fuel design. Extended
burnup of the M5 material has not been shown to alter the functions
of the rod cladding, which is to provide a barrier against the
release of radioactive material. Initial plant conditions, which are
considered in the accident analysis, will also be maintained such
that no new plant conditions will exist that could affect the
analysis results. Since plant functions and conditions are not
impacted by the proposed revision and the higher burnup limit of the
Mark-BW fuel design with M5 alloy material is not postulated to
become an accident initiator based on the similarity with Mark-B
fuel design and Zircaloy-4 material, the possibility of a new or
different kind of accident is not created.
The proposed changes will not alter the plant configuration or
require any new or unusual operator actions. They do not alter the
way any structure, system, or component functions and do not alter
the manner in which the plant is operated. These changes do not
introduce any new failure modes.
Therefore, the proposed change does 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 margin of safety is established by the acceptance criteria
used by NRC. Meeting the acceptance criteria assures that the
consequences of accidents are within known and acceptable limits.
The emergency core cooling system (ECCS) acceptance criteria are not
exceeded. Testing has been performed on M5 alloy with respect to
criteria for peak cladding temperature (PCT) and maximum cladding
oxidation. These tests demonstrate that M5 alloy rod cladding
remains within PCT of 2200 degrees Fahrenheit and conservatively
bounded by the 17 percent limit for maximum cladding oxidation. M5
alloy oxidation rates are lower than that of Zircaloy at
temperatures less than 2200 degrees Fahrenheit and have similar
rates for temperatures up to about 2300 degrees Fahrenheit. High-
temperature oxidation rates of M5 alloy remain equivalent to
Zircaloy and, as such, respond as hydrogen generators to the same
extent. Core geometry for amenable cooling is not directly related
to rod cladding material; however, it applies equally well to all
materials. The consequences of both thermal and mechanical
deformation of fuel assemblies have been assessed, and the resultant
deformations have been shown to maintain coolable core
configurations. The ECCS is evaluated against the thermal power
immediately after shutdown. The thermal power is largely a function
of short-lived fission products which tend to saturate at relatively
low burnup limits and are not appreciably affected by extended
burnup. Therefore, with no system changes being proposed; long-term
cooling is maintained. Additionally, the fuel storage cooling system
is capable of supporting the long-term storage of the extended
burnup fuel assemblies' decay heat.
The changes to burnup limit have been evaluated against
Departure from Nucleate Boiling (DNB) events and all applicable
acceptance criteria are met. In addition, the proposed revision to
allow an increase in the burnup limit of the Mark-BW fuel design
with M5 alloy will not impact plant setpoints that maintain the
margin of safety. Based on these results, it is concluded that the
margin of safety is not significantly reduced. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety.
Removing revision numbers, dates, and parenthetical information
from the listed TRs will not reduce a margin of safety because this
information has no effect on any safety analysis assumption nor does
it revise any setpoints assumed in the analysis of record. The
proposed change is consistent with NUREG-1431, issued by the NRC
staff, revising the TSs to reflect the approved level of detail,
which indicates that there is no 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Michael L. Marshall, Jr.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: May 25, 2006 (TSC 06-02).
Description of amendment request: The proposed amendment would
revise Section 6.2.1.6 of the Sequoyah Nuclear Plant (SQN) Updated
Final Safety Analysis Report (UFSAR). This change would revise the
methodology used for containment sump debris transport analysis and
affects SQN's current design and licensing basis described in Section
6.2.1.6 of the SQN UFSAR.
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 design function of the sump during accident conditions is to
support emergency core cooling systems (ECCS) and containment spray
system operation for
[[Page 35461]]
recirculation. The sump is a passive feature that does not act as an
accident initiator, (i.e., failure of the sump would not initiate a
design basis accident).
The proposed change to the UFSAR regarding debris transport
analysis provides an overall improvement in the analysis for
recirculation operation and does not change the consequences of
accidents previously evaluated. The change in methodology is neutral
with regard to probability. Consequently, the changes associated
with the enclosed license amendment do not affect the frequency of
occurrence for accidents previously evaluated in the UFSAR.
Accident dose as previously evaluated in the UFSAR is unaffected
by the proposed license amendment.
Based on the above discussion, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The sump is a passive component and is not an accident
initiator; i.e., failure of the sump will not initiate a design
basis accident. The sump transport methodology is used to confirm
the ability of the sump to perform all safety functions during
normal and accident conditions. Consequently, this activity does not
create a possibility of a new or different type of accident than any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The changes addressed in TVA's proposed amendment are associated
with methodology for debris transport to the containment sump.
The change does not affect specific safety limits, design
limits, set points, or other critical parameters. The transport
methodology is used to confirm that the ECCS and containment spray
systems will perform their safety functions for all accident
conditions within existing equipment performance capability margins.
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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Michael L. Marshall, Jr.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: May 9, 2006.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) 1.1, ``Definitions,'' and 3.4.16,
``RCS [reactor coolant system] Specific Activity.'' The revisions would
replace the current Limiting Condition for Operation (LCO) 3.4.16 limit
on RCS gross specific activity with limits on RCS Dose Equivalent I-131
and Dose Equivalent Xe-133 (DEX). The conditions and required actions
for LCO 3.4.16 not being met, and surveillance requirements for LCO
3.4.16, are being revised. The modes of applicability for LCO 3.4.16
would be extended. The current definition of [Emacr]--Average
Disintegration Energy in TS 1.1 would be replaced by the definition of
DEX. In addition, the current definition of Dose Equivalent I-131 in TS
1.1 would be revised to allow alternate, NRC-approved thyroid dose
conversion factors.
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 change[s] involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes would add new thyroid dose conversion
factor reference[s] to the definition of DOSE EQUIVALENT I-131,
eliminate the definition of [Emacr]--AVERAGE DISINTEGRATION ENERGY,
add a new definition of DOSE EQUIVALENT XE-133, replace the
Technical Specification (TS) 3.4.16 limit on reactor coolant system
(RCS) gross specific activity with a limit on noble gas specific
activity in the form of a Limiting Condition for Operation (LCO) on
DOSE EQUIVALENT XE-133, increase the Completion Time for Required
Action B.1, replace TS Figure 3.4.16-1 with a maximum limit on DOSE
EQUIVALENT I-131, extend the Applicability of LCO 3.4.16, and make
corresponding changes to TS 3.4.16 to reflect all of the above. The
proposed changes are not accident initiators and have no impact on
the probability of occurrence of any design basis accidents.
The proposed changes will have no impact on the consequences of
a design basis accident because they will limit the RCS noble gas
specific activity to be consistent with the values assumed in the
radiological consequence analyses. The changes will also limit the
potential RCS [radio]iodine concentration excursion to the value
currently associated with full power operation, which is more
restrictive on plant operation than the existing allowable RCS
[radio]iodine specific activity at lower power levels.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. [Do] the proposed change[s] create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not alter any physical part of the plant
nor do they affect any plant operating parameters besides the
allowable specific activity in the RCS. The changes which impact the
allowable specific activity in the RCS are consistent with the
assumptions assumed in the current radiological consequence
analyses. [The proposed changes are also not accident initiators.]
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 change[s] involve a significant reduction
in a margin of safety?
Response: No.
The acceptance criteria related to the proposed changes involve
the allowable control room and offsite radiological consequences
following a design basis accident. The proposed changes will have no
impact on the radiological consequences of a design basis accident
because they will limit the RCS noble gas specific activity to be
consistent with the values assumed in the radiological consequence
analyses. The changes will also limit the potential RCS
[radio]iodine specific activity excursion to the value currently
associated with full power operation, which is more restrictive on
plant operation than the existing allowable RCS [radio]iodine
specific activity at lower power levels.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: David Terao.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: May 11, 2006.
Description of amendment request: The proposed amendment would
revise Surveillance Requirements 3.7.2.1, 3.7.3.1, and 3.7.3.3 on
verifying the
[[Page 35462]]
closure time of the main steam isolation valves (MSIVs), main feedwater
regulating valves (MFRVs), main feedwater regulating valve bypass
valves (MFRVBVs), and main feedwater isolation valves (MFIVs) in the
Technical Specifications (TSs). These valves are the Main Steam and
Main Feedwater System isolation valves. The revisions would replace (1)
the specified maximum acceptable valve closure time for the MSIVs,
MFRVs, and MFRVBVs, and (2) TS Figure 3.7.3-1, which shows acceptable
valve closure times for the MFIVs, by the reference to the valve
closure time, is verified to be ``within limits.'' The maximum
acceptable valve closure times for the MFRVs and MFRVBVs, and TS Figure
3.7.3-1 will be relocated to the TS Bases. The maximum acceptable valve
closure time for the MSIV is already in the TS Bases.
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.
Because the proposed change[s remove] specific isolation times
from the TS and [relocate] the specific values to the TS Bases,
there are no design or physical changes to the facility or to the
Main Steam and Main Feedwater System isolation valves themselves.
The design and functional performance requirements, operational
characteristics, and reliability of these components remain
unchanged. There is[,] therefore[,] no impact on the design safety
function of the valves to close (as an accident mitigator), nor is
there any change with respect to inadvertent closure (as a potential
transient initiator). Since no failure mode or initiating condition
that could cause an accident (including any plant transient)
evaluated per the FSAR [Final Safety Analysis Report]-described
safety analyses is created or affected, the change cannot involve a
significant increase in the probability of an accident previously
evaluated. The probability of an accident is not affected. The Main
Steam and Main Feedwater System isolation valves are assumed to
function to mitigate some accidents (for example, SLB [steam line
break] and FWLB [main feedwater line break]). The proposed change[s]
only [affect] the level of detail included in the TS. The TS
requirements continue to provide the same level of assurance as
before that the Main Steam and Main Feedwater System isolation
valves are capable of performing their intended safety function.
These isolation valves will continue to be verified operable in the
same manner as before. As such, the proposed change[s do] not affect
the ability of the isolation valves to perform their assumed
mitigation function.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an 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.
The proposed change[s] only [affect] the level of detail
included in the TS. The TS requirements [are not being changed and
they will] continue to provide the same level of assurance as before
that the Main Steam and Main Feedwater System isolation valves are
capable of performing their intended safety function. The Main Steam
and Main Feedwater System isolation valves will continue to be
verified operable in the same manner. As such, the proposed change[s
do] not involve a modification to the physical configuration of the
plant (i.e., no new equipment will be installed) or change in the
methods governing normal plant operation. The proposed change[s]
will not impose any new or different requirements or introduce a new
accident initiator, accident precursor, or malfunction mechanism.
Additionally, there is no change in the types or increases in the
amounts of any effluent that may be released off-site and there is
no increase in individual or cumulative occupational exposure.
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 change[s] involve a significant reduction
in a margin of safety?
Response: No.
The proposed change[s do] not reduce the margin of safety. The
proposed change[s] only [affect] the level of detail included in the
TS. The TS requirements [are not being changed and will] continue to
provide the same level of assurance as before that the Main Steam
and Main Feedwater System isolation valves will continue to be
verified operable in the same manner as before. As such, the
proposed change[s do] not affect the assumptions of any accident
analysis or the availability or operability of any plant equipment.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: David Terao.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: February 14, 2006.
Brief description of amendment request: The proposed amendments
would add a requirement to the Title 10 of the Code of Federal
Regulations, (10 CFR) part 50 license to restrict the minimum cooling
time and burnup of spent fuel assemblies that will be placed into
storage in the NUHOMS HD spent fuel dry storage system at Surry
starting in the summer of 2006.
Date of publication of individual notice in Federal Register: May
16, 2006 (71 FR 28390).
Expiration date of individual notice: 30 day expiration date, June
15, 2006, and 60 day expiration date, July 17, 2006.
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
[[Page 35463]]
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].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of application for amendment: April 26, 2004, as supplemented
April 18 and October 11, 2005, and May 19, 2006.
Brief description of amendment: The amendment revised Technical
Specification 3.8.7, ``Inverters--Operating'' to change the completion
time for restoration of an inoperable Division 1 or 2 inverter from the
current 24 hours to 7 days.
Date of issuance: May 26, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days of the date of issuance.
Amendment No.: 174.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: June 8, 2004 (69 FR
32072). The supplements dated April 18 and October 11, 2005, and May
19, 2006, provided additional information that clarified the
application, but did not expand the scope of the application as
originally noticed, and did not change the NRC staff's original
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 26, 2006.
No significant hazards consideration comments received: No.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: June 3, 2005, as supplemented
by letter dated March 7, 2006.
Brief description of amendments: The amendments revise the Updated
Final Safety Analysis Report (UFSAR) to incorporate the description of
the approved changes associated with the plant modifications made to
the diesel generator cooling water system for each emergency diesel
generator as described in the amendment application of June 3, 2005, as
supplemented by letter dated March 7, 2006.
Date of issuance: May 25, 2006.
Effective date: As of the date of issuance to be implemented within
90 days from the date of issuance.
Amendment Nos.: Unit 1-160, Unit 2--160, Unit 3 -160.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revise the Operating Licenses and the UFSAR for all three
units.
Date of initial notice in Federal Register: July 5, 2005 (70 FR
38715). The March 7, 2006, supplemental letter provided additional
clarifying information, did not expand the scope of the application as
originally noticed, and did not change the NRC staff's original
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 25, 2006.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: August 20, 2004, supplemented
January 31, 2006.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3.3.8, ``Post Accident Monitoring (PAM)
Instrumentation,'' to eliminate TS requirements associated with the
reactor building spray flow instruments commensurate with the
importance of their post-accident function.
Date of Issuance: June 1, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 350/352/351.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Licenses and the Technical Specifications.
Date of initial notice in Federal Register: September 28, 2004 (69
FR 57983).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 1, 2006.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
2, Pope County, Arkansas
Date of amendment request: September 19, 2005.
Brief description of amendment: The proposed changes would revise
Technical Specification (TS) 3.1.1.5, ``Minimum Temperature for
Criticality.'' The request proposes to change the current Limiting
Condition for Operation (LCO) for TS 3.1.1.5 by raising the minimum
temperature for criticality from the current value of >= 525 [deg]F to
>= 540 [deg]F; to change the current Action statement for LCO 3.1.1.5
to reflect this change; and to delete the current statement in
Surveillance Requirement 4.1.1.5 and replace the statement with wording
consistent with NUREG-1432, ``Standard Technical Specifications
Combustion Engineering Plants.'' Also, changes will be made to the ANO-
2 TS Bases in accordance with the Technical Specifications (TS) Bases
Control Program (ANO-2 TS 6.5.14).
Date of issuance: May 30, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 264.
Renewed Facility Operating License No. NPF-6: The amendment revised
the Technical Specifications and Surveillance Requirements.
Date of initial notice in Federal Register: December 6, 2005, (70
FR 72672).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 30, 2006.
No significant hazards consideration comments received: No.
[[Page 35464]]
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: January 20, 2005, as
supplemented July 5, 2005.
Brief description of amendments: The amendments revised several
Technical Specifications (TSs) using six TS Task Force (TSTF) generic
changes. The six TSTFs (nos. 5, 93, 258, 299, 308, and 361) delete
redundant safety limit violation notification requirements; extend the
pressurizer heater surveillance frequency from 92 days to 18 months;
remove redundant requirements and add other requirements to the
Administrative Controls section of the TSs; clarify the requirements
regarding the frequency of testing for cumulative and projected dose
contributions from radioactive effluents; and add a note to the
residual heat removal requirements during Mode 6 low water level
operations that allows one required residual heat removal (RHR) loop to
be inoperable for up to 2 hours for surveillance testing provided the
other RHR loop is operable and in operation.
The amendments represent partial approval of the January 20, 2005,
application for the proposed amendments. The Commission has granted the
request of Florida Power and Light Company (the licensee) to withdraw
portions of its January 20, 2005, application for the proposed
amendment. The application also included TSTF-95, which would extend
the completion time for reducing the Power Range High trip setpoint
from 8 to 72 hours and TSTF-101, which would change the auxiliary
feedwater pump test frequency to be consistent with the inservice test
program frequency. However, by letter dated March 22, 2005, the
licensee withdrew the request to adopt TSTF-95 and by letter dated
October 13, 2005, the licensee withdrew the request to adopt TSTF-101.
Date of issuance: May 26, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos: 229 and 225.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the TSs.
Date of initial notice in Federal Register: March 15, 2005 (70 FR
12747). The supplement dated July 5, 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 amendments is contained
in a Safety Evaluation dated May 26, 2006.
No significant hazards consideration comments received: No.
Nuclear Management Company, Docket No. 50-263, Monticello Nuclear
Generating Plant (MNGP), Wright County, Minnesota
Date of application for amendment: June 29, 2005, as supplemented
by letters dated April 25 (two letters), May 4, and May 12, 2006.
Brief description of amendment: The amendment converts the current
Technical Specifications (CTSs) to the Improved Technical
Specifications (ITSs) format and relocates certain requirements to
other licensee-controlled documents. The ITSs are based on NUREG-1433,
``Standard Technical Specifications General Electric Plants BWR/4,''
Revision 3, dated June 2004; the Commission's Final Policy Statement,
``NRC Final Policy Statement on Technical Specification Improvements
for Nuclear Power Reactors,'' dated July 22, 1993 (58 FR 39132); and 10
CFR 50.36, ``Technical specifications.'' The purpose of the conversion
is to provide clearer and more readily understandable requirements in
the TSs for MNGP to ensure safer operation of the unit. In addition,
the amendment includes a number of issues that are considered beyond
the scope of NUREG-1433.
Date of issuance: June 5, 2006.
Effective date: As of the date of issuance and shall be implemented
by September 30, 2006.
Amendment No: 146.
Facility Operating License No. DPR-22: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: November 16, 2005 (70
FR 70889).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 5, 2006.
No significant hazards consideration comments received: No.
Amendment No: 146.
Facility Operating License No. DPR-22: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: November 16, 2005 (70
FR 70889).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 5, 2006.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: October 19, 2005, as
supplemented by letter dated December 23, 2005.
Brief description of amendments: The amendments updated the
Technical Specification (TS)5.3, ``Unit Staff Qualifications,''
operator minimum qualification requirements contained in the March 28,
1980, NRC letter to all licensees with the more recent NRC-approved
operator qualification requirements contained in American National
Standards Institute/American Nuclear Society (ANSI/ANS) 3.1-1993. In
addition, the changes removed the TS 5.3.1 plant staff retraining and
replacement training program requirements, which have been superseded
by requirements contained in 10 CFR 50.120.
Date of issuance: May 26, 2006.
Effective date: As of its date of issuance, and shall be
implemented within 90 days of issuance.
Amendment Nos.: Unit 1--187 ; Unit 2--189.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 20, 2005 (70
FR 75495). The December 23, 2005, supplemental letter provided
additional information that clarified the application, and did not
expand the scope of the application as originally noticed.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 26, 2006.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket No. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of application for amendments: February 28, 2006, as
supplemented on April 7, 2006.
Brief description of amendments: The amendments revise the SSES 1
and 2 Technical Specification (TS) Surveillance Requirements 3.8.4.7
and 3.8.4.8 to clarify that Diesel Generator ``E'' (DG E) electrical
power subsystem testing does not require a mode restriction when the DG
E diesel is not aligned to the Class 1E distribution system.
Date of issuance: May 30, 2006.
[[Page 35465]]
Effective date: As of the date of issuance and to be implemented
within 30 days.
Amendment Nos.: 235 and 212.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the TSs and license.
Date of initial notice in Federal Register: March 28, 2006 (71 FR
15485). The supplement dated April 7, 2006, 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.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 30, 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 10, 2005, as
supplemented by letters dated July 14, 2005, and October 20, 2005.
Brief description of amendments: The amendments modified Technical
Specification Surveillance Requirement 4.5.3.2 b to allow safety
injection and charging pumps to run in a recirculation flow path,
provided that two independent means are used to prevent injection into
the reactor coolant system.
Date of issuance: May 31, 2006.
Effective date: As of the date of issuance, and shall be
implemented in 60 days.
Amendment Nos.: 273 and 254.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 12, 2005 (70 FR
19116). The supplements dated July 14, 2005 and October 20, 2005
provided clarifying information only and did not change the initial no
significant hazards consideration determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated May 31, 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: August 31, 2005, as
supplemented by letters dated December 8, 2005, and April 10, 2006.
Brief description of amendments: The amendments changed the
Technical Specifications (TSs) to move the requirements for the
containment area high-range radiation monitors from TS 3/4.3.3.1,
``Radiation Monitoring Instrumentation,'' to TS 3/4.3.3.7, ``Accident
Monitoring Instrumentation,'' and correct a typographical error in
Surveillance Requirement 4.2.2.
Date of issuance: May 25, 2006.
Effective date: May 25, 2006.
Amendment Nos.: 272 and 253.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the TSs.
Date of initial notice in Federal Register: January 17, 2006 (71 FR
2594). The April 10, 2006 supplement did not expand the scope of the
application, as originally noticed, and did not change the staff's
original proposed no significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 25, 2006.
No significant hazards consideration comments received: No.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: April 29, 2005, as supplemented
on August 15 and December 9, 2005, and January 11 and 25, and May 9,
2006.
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.5.1, ``Accumulators,'' and TS 3.5.4, ``Refueling
Water Storage Tank,'' to reflect the results of revised analyses
performed to accommodate the proposed extended power uprate and revises
TS 5.6.4, ``Core Operating Limits Report,'' to permit the use of
approved methodology for large-break and small-break loss-of-coolant
accident analyses.
Date of issuance: May 31, 2006.
Effective date: As of the date of issuance to be implemented prior
to restart from the fall 2006 refueling outage.
Amendment No.: 96.
Renewed Facility Operating License No. DPR-18: Amendment revised
the Technical Specifications and the license.
Date of initial notice in Federal Register: June 7, 2005 (70 FR
33219). The August 15 and December 9, 2005, and January 11 and 25, and
May 9, 2006, letters 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 May 31, 2006.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: March 17, 2006, as supplemented
on April 14, 2006. The supplemental letter dated April 14, 2006,
provided clarifying information that did not change the scope of the
March 17, 2006, application nor the initial proposed no significant
hazards consideration determination.
Brief description of amendments: The amendments authorized the
licensee to credit administering potassium iodide (KI) to reduce the
30-day post-accident thyroid dose to the occupants of the main control
room for an interim period of 4 years. In addition, the design-basis
accident analysis section of the Updated Final Safety Analysis Reports
will be updated to reflect crediting of KI.
Date of issuance: May 25, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 249 and 193.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Operating Licenses.
Date of initial notice in Federal Register: March 27, 2006 (71 FR
15223). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 25, 2006.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: July 21, 2005.
Brief Description of amendments: These amendments revised the
Technical Specifications (TSs) to change the accident monitoring
instrumentation listing, allowed outage times, requirements, and
surveillances to be consistent with the requirements of the Improved
TSs for post-accident monitoring instrumentation.
Date of issuance: May 31, 2006.
[[Page 35466]]
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 247/246.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments change the Technical Specifications.
Date of initial notice in Federal Register: January 3, 2006 (71 FR
155).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 31, 2006.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this June 12, 2006.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E6-9434 Filed 6-19-06; 8:45 am]
BILLING CODE 7590-01-P