[Federal Register Volume 71, Number 99 (Tuesday, May 23, 2006)]
[Notices]
[Pages 29671-29686]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-4736]
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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 April 28, 2006 to May 11, 2006. The last
biweekly notice was published on May 9, 2006 (71 FR 26995).
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
[[Page 29672]]
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, 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-
[[Page 29673]]
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].
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of amendment request: April 6, 2006.
Description of amendment request: The proposed amendment would
allow the use of a different methodology for determining the design
requirements necessary for protecting safety-related equipment from
damage by tornado generated missiles.
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 amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The probability of occurrence of an accident previously
evaluated is not significantly increased by the proposed change to
permit probabilistic evaluation of missiles generated by natural
phenomena. The actual frequency of tornado occurrence at Kewaunee is
unaffected by the proposed change in assessment methodology.
Furthermore, the projected frequency of tornado occurrence, as
specified in the USAR [Updated Safety Analysis Report], is not
significantly affected by this change. The value for the probability
of tornado occurrence in the updated study is in general agreement
with the original value in the USAR (i.e. 3.97E-4 vs. 4.86E-4).
Similarly, the probability of a tornado-generated missile is not
significantly affected by this change.
Likewise, the consequences of an accident previously evaluated
are not significantly increased by the proposed change. The actual
probability of a tornado missile onsite remains unchanged. The
actual probability of a tornado missile strike remains unchanged.
For the limited number of components affected by this proposed
change (i.e. exhaust ducts and fuel vent), the missile strike
probability is approximately 5.75 x 10-7 per year, which
is significantly lower than the SRP [Standard Review Plan]
acceptance criteria of 1 x 10-6 per year. Therefore, the
proposed change is not considered to constitute a significant
increase in the consequences of an accident due to the low
probability of occurrence.
In addition, use of a probabilistic versus a deterministic
methodology to assess missile hazard acceptability has no impact on
accident initiation or consequence. Therefore, the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes involve use of an evaluation methodology to
determine protection requirements for two specific support
components for safety-related equipment, which may be adversely
affected by missiles during a tornado. A tornado at Kewaunee is
considered in the USAR as a separate event and not occurring
coincident with any of the design basis accidents in the USAR. As
such, no new or different kind of accident is created by the
proposed change to permit probabilistic evaluation of missiles
generated by natural phenomena.
This change involves recognition of the acceptability of
performing tornado missile strike probability calculations in
accordance with established regulatory guidance in lieu of using
deterministic methodology alone. Therefore, the change would not
create the possibility of, or be the initiator for, any new or
different kind of accident. The acceptance criterion of the SRP
guidance establishes a threshold for tornado missile damage to
system components that is consistent with this conclusion.
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 amendment involve a significant reduction
in a margin of safety?
Response: No.
The request does not involve a significant reduction in a margin
of safety. The existing design basis for Kewaunee, with respect to a
tornado affecting safety related equipment, is to provide positive
missile barriers for all safety-related systems and components. The
proposed change recognizes that for probability of occurrences below
the SRP established acceptance limit, the extremely low probability
associated with an ``important'' system or component being struck by
a tornado missile does not represent a significant reduction in the
margin of safety provided by use of the deterministic methodology.
The change from ``protecting all safety-related systems and
components'' to ``an extremely low probability of occurrence of
tornado generated missile strikes on portions of important systems
and components'' is not considered to constitute a significant
reduction in the margin of safety.
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: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Branch Chief: L. Raghavan.
Dominion Nuclear Connecticut, Inc., Docket No. 50-423 Millstone Power
Station, Unit No. 3 New London County, Connecticut
Date of amendment request: March 28, 2006.
Description of amendment request: The proposed amendment would
eliminate redundant surveillance requirements [SRs] pertaining to post-
maintenance/post-modification testing. The associated TS bases will be
updated to address the proposed changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes do not modify any plant equipment and do
not impact any failure modes that could lead to an accident. Testing
in accordance with the requirements of SR 4.0.1 will continue to
provide the necessary assurance that the associated systems will
function consistent with the assumptions used in the accident
analyses. On this basis, the proposed amendment does not increase
the probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not involve any physical changes to
systems, structures, or components, or involve a change to the
[[Page 29674]]
method of plant operation. The requirement to perform post
maintenance/post modification testing will continue to be
implemented consistent with SR 4.0.1, through existing plant
programs and procedures. As such, the proposed amendment does not
introduce any new failure modes, accident initiators or malfunctions
that would cause a new or different kind of accident. Therefore, the
proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The TS changes do not involve a significant reduction in a
margin of safety because the requirements described in SR 4.0.1, as
implemented through existing plant programs and procedures, will
continue to ensure that post maintenance/post modification testing
will be performed when necessary. The proposed change does not
affect any of the assumptions used in the accident analyses, nor
does it affect operability requirements for equipment important to
plant safety. Therefore, the margin of safety is not impacted by the
proposed amendment.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Branch Chief: Darrell J. Roberts.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: April 17, 2006.
Description of amendment request: The proposed amendment would
change the method for calculating fuel pool decay heat load from the
original licensing basis methodology of ORIGEN and the Auxiliary
Systems Branch Technical Position (ASBTP) 9-2, ``Residual Decay Heat
Energy for Light Water Reactors for Long-Term Cooling,'' to ORIGEN-ARP.
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 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 adoption of ORIGEN-ARP does not affect the probability or
consequences of an accident previously evaluated. The calculation of
the fuel pool decay heat load is used to evaluate and demonstrate
the ability of the fuel pool cooling system to maintain the fuel
pool temperatures within the acceptance limits specified in the
Columbia Final Safety Analysis Report [FSAR]. The proposed change to
the methodology used to calculate the fuel pool [decay] heat load
has no bearing on the probability or consequences of any previously
evaluated accident. Therefore, this 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 change involves the use of a different methodology for
calculating fuel pool decay heat load. This change does not involve
any new equipment, it does not change any previously approved
acceptance limits, and it does not affect or alter the operation of
any equipment. Therefore[,] this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety provided by the fuel pool cooling system is
primarily defined by the difference between the maximum allowed fuel
pool temperature and the boiling point of water. The margin of
safety is supplemented by the ability to make up water to the spent
fuel pool if boiling were to occur. The proposed change in
methodology for calculating the fuel pool [decay] heat load does not
alter the current temperature limits or acceptance criteria
specified in the FSAR and has no effect on the ability to provide
make-up water if boiling were to occur. This change will allow
Energy Northwest to more accurately calculate the fuel pool [decay]
heat load to provide added confidence in the ability of the fuel
pool cooling system to accommodate the heat load added to the spent
fuel pool during refueling activities. 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: William A. Horin, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: April 18, 2006.
Description of amendment request: The proposed change would modify
technical specification surveillance requirement 3.6.1.1.2 by changing
the test frequency of the drywell-to-suppression chamber bypass leakage
test from 24 to 120 months. This proposed amendment also includes
testing the suppression chamber-to-drywell vacuum breakers on a 24-
month frequency.
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 operation of Columbia Generating Station in
accordance with the proposed amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response: No.
The proposed changes would modify Technical Specification (TS)
Surveillance Requirement (SR) 3.6.1.1.2 and add two new SRs, SR
3.6.1.1.3 and SR 3.6.1.1.4. The proposed changes will extend the
frequency for the drywell-to-suppression chamber bypass leakage test
while maintaining the current leakage testing frequency for the
suppression chamber-to-drywell vacuum breakers, and establish
leakage acceptance criteria for the suppression chamber-to-drywell
vacuum breakers when the valves are tested individually.
The performance of a drywell-to-suppression chamber bypass
leakage test or suppression chamber-to-drywell vacuum breaker
leakage test is not a precursor to any accident previously
evaluated. Thus, the proposed changes to the performance of the
leakage tests do not have any affect on the probability of an
accident previously evaluated.
The performance of a drywell-to-suppression chamber bypass
leakage test or a suppression chamber-to-drywell vacuum breaker
leakage test continues to provide assurance that the containment
will perform as designed. Thus, the radiological consequences of any
accident previously evaluated are not impacted.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the operation of Columbia Generating Station in
accordance with the proposed amendment create the possibility of a
new or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes to TS SR 3.6.1.1.2, and the addition of SR
3.6.1.1.3, and SR 3.6.1.1.4 do not affect the assumed performance of
any Columbia Generating
[[Page 29675]]
Station structure, system or component previously evaluated. The
proposed changes do not introduce any new modes of system operation
or any new failure mechanisms. This is an administrative change and
does not involve the modification, addition or removal of any plant
equipment.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the operation of Columbia Generating Station in
accordance with the proposed amendment involve a significant
reduction in the margin of safety?
Response: No.
The current frequency associated with a drywell-to-suppression
chamber bypass leakage test in TS SR 3.6.1.1.2 is 24 months or 12
months if two consecutive tests fail and continues at this frequency
until two consecutive tests pass. The proposed change will modify
this leakage test frequency to 120 months, or 48 months following
one test failure or 24 months if two consecutive tests fail and
continues at this frequency until two consecutive tests pass. The
proposed change in SR 3.6.1.1.2 frequency is acceptable as the
results from previous tests show that the measured drywell-to-
suppression chamber bypass leakage at the current TS frequency has
been a small percentage of the allowable leakage. Acceptability is
further demonstrated by the design requirements applied to the
primary containment components and other periodically performed
primary containment inspections.
The proposed SR 3.6.1.1.3 will establish a leakage test
frequency of 24 months for each suppression chamber-to-drywell
vacuum breaker except when the leakage test of SR 3.6.1.1.2 has been
performed within the past 24 months. SR 3.6.1.1.3 specifies a
leakage limit for each suppression chamber-to-drywell vacuum breaker
pathway of less than or equal to 12 percent of the bypass leakage
limit of SR 3.6.1.1.2. The proposed SR 3.6.1.1.4 will establish a
total leakage limit of less than or equal to 30 percent of the
bypass leakage limit of SR 3.6.1.1.2 when the suppression chamber-
to-drywell vacuum breakers are tested in accordance with SR
3.6.1.1.3.
TS SR 3.6.1.1.2 drywell-to-suppression chamber bypass leakage
test monitors the combined leakage of three types of pathways: (1)
The drywell floor and downcomers, (2) piping externally connected to
both the drywell and suppression chamber air space, and (3) the
suppression chamber-to-drywell vacuum breakers. This amendment would
extend the surveillance interval on the passive components of the
test (the first two types of pathways), while retaining the current
surveillance interval on the active components (suppression chamber-
to-drywell vacuum breakers). The proposed changes establish leakage
limits for both individual suppression chamber-to-drywell vacuum
breakers and the total leakage. Additional testing is required if
acceptable results are not achieved.
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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: April 12, 2006.
Description of amendment request: The proposed amendment would
revise the Technical Specification reactor pressure vessel Pressure and
Temperature (P-T) curves.
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 License Amendment (LA) does not involve a
significant increase in the probability or consequences of an
accident previously evaluated. There are no physical changes to the
plant being introduced by the proposed changes to the pressure-
temperature curves. The proposed change does not modify the reactor
coolant pressure boundary, (i.e., there are no changes in operating
pressure, materials, or seismic loading). The proposed change does
not adversely affect the integrity of the reactor coolant pressure
boundary such that its function in the control of radiological
consequences is affected.
The proposed pressure-temperature curves are generated in
accordance with the fracture toughness requirements of 10 CFR 50
Appendix G, and American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel (B&PV) Code, Section Xl, Appendix G and
Regulatory Guide (R.G.) 1.99, Revision 2[,] ``Radiation
Embrittlement of Reactor Vessel Materials.'' A best-estimate
calculation of reactor vessel 34 effective full power years (EFPYs)
neutron fluence and associated uncertainty has been completed for
Pilgrim using the Radiation Analysis Modeling Application (RAMA)
methodology. This methodology was previously approved by the NRC.
The resulting reactor vessel neutron fluence value was then used in
conjunction with R.G. 1.99, [Revision] 2 to determine the adjusted
reference temperature (ART) and with ASME Section Xl Appendix G to
develop revised P-T curves.
This provides sufficient assurance that the Pilgrim reactor
vessel will be operated in a manner that will protect it from
brittle fracture under all operating conditions. This proposed
license amendment provides compliance with the intent of 10 CFR
[Part 50] Appendix G and provides margins of safety that assure
reactor vessel integrity.
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 license amendment does not create the possibility
of new or different kind of accident from any accident previously
evaluated. The revised pressure-temperature curves are generated in
accordance with the fracture toughness requirements of 10 CFR Part
50 Appendix G and ASME Section Xl Appendix G. Compliance with the
proposed pressure-temperature curves will ensure the avoidance of
conditions in which brittle fracture of primary coolant pressure
boundary materials is possible because such compliance with the
pressure-temperature curves provides sufficient protection against a
non-ductile-type fracture of the reactor pressure vessel. No new
modes of operation are introduced by the proposed change. The
proposed change will not create any failure mode not bounded by
previously evaluated accidents. Further, the proposed change does
not affect any activities or equipment and is not assumed in any
safety analysis to initiate any accident sequence. This provides
sufficient assurance that Pilgrim reactor vessel will be operated in
a manner that will protect it from brittle fracture under all
operating conditions.
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 proposed license amendment requests the use of revised P-T
curves that are based on established NRC and ASME methodologies. A
best-estimate calculation of reactor vessel neutron fluence and
associated uncertainty has been completed for Pilgrim through 34
EFPY using the NRC approved RAMA methodology. The 34 EFPY reactor
vessel neutron fluence value was used in conjunction with R.G. 1.99,
[Revision 2] to compute reference temperature shift, and with ASME
Section Xl Appendix G to develop revised P-T curves. This provides
sufficient margin such that the Pilgrim reactor vessel will be
operated in a manner that will protect it from brittle fracture
under all operating conditions. Operation within the proposed limits
ensures that the reactor vessel materials will continue to behave in
a non-brittle manner, thereby preserving the original safety design
bases. No plant safetylimits, set points, or design parameters are
adversely affected by the proposed changes.
[[Page 29676]]
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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Travis C. McCullough, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
Branch Chief: Richard Laufer.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: June 2, 2005.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) reactor coolant system leakage
detection instrumentation requirements and actions.
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 relocation is administrative in
nature and does not involve the modification of any plant equipment
or affect basic plant operation. The associated instrumentation and
surveillances are not assumed to be an initiator of any analyzed
event, nor are these functions assumed in the mitigation of
consequences of accidents. Additionally, the associated required
actions for inoperable components do not impact the initiation or
mitigation of any accident. 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 change does not involve any physical
alteration of plant equipment and does not change the method by
which any safety-related system performs its function. As such, no
new or different types of equipment will be installed, and the basic
operation of installed equipment is unchanged. The methods governing
plant operation and testing remain consistent with current safety
analysis assumptions. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The proposed change to relocate current TS
requirements to the FSAR [Final Safety Analysis Report], consistent
with regulatory guidance and previously approved changes for other
stations, are administrative in nature. These changes do not negate
any existing requirement, and do not adversely affect existing plant
safety margins or the reliability of the equipment assumed to
operate in the safety analysis. As such, there are no changes being
made to safety analysis assumptions, safety limits or safety system
settings that would adversely affect plant safety as a result of the
proposed change. Margins of safety are unaffected by requirements
that are retained, but relocated from the Technical Specifications
to the FSAR. Additionally, the changes being made to allow
additional repair time for inoperable instrumentation will not
affect the required leakage limits, which will continue to be
monitored at the same required frequency. These compensatory
measures, operational limitations, and administrative functions that
will be modified are not credited in any design-basis event and do
not reflect a margin of safety. 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Travis C. McCullough, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
Branch Chief: Richard Laufer.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos.
1 and 2, Will County, Illinois.
Date of amendment request: November 18, 2005.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) to adopt NRC-approved Revision
4 to Technical Specification Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-449, ``Steam Generator Tube
Integrity.'' The proposed amendment would also include changes to the
TS definition of Leakage, TS 3.4.13, ``RCS [Reactor Coolant System]
Operational LEAKAGE,'' TS 5.5.9, ``Steam Generator (SG) Program,'' TS
5.6.9, Steam Generator Tube Inspection Report,'' and would add TS
3.4.19, ``Steam Generator (SG) Tube Integrity.'' The proposed changes
are necessary in order to implement the guidance for the industry
initiative on Nuclear Energy Institute (NEI) 97-06, ``Steam Generator
Program Guidelines.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments
adopting TSTF-449, including a model safety evaluation and model no
significant hazards consideration (NSHC) determination, using the
consolidated line item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on May 6, 2005
(70 FR 24126). The licensee affirmed the applicability of the published
NSHC determination in its application dated November 18, 2005.
The licensee included a variation from TSTF-449 for Braidwood, Unit
2 and Byron, Unit 2 in that the proposed amendment would also include
an effective change to the definition of primary pressure boundary from
the hot-leg tube end weld to 17 inches below the top of the hot-leg
tube sheet. The proposed amendment would also delete the current TS
allowance to use Westinghouse laser welded sleeves as a SG tube repair
method. The licensee provided an analyses of the NSHC issue in its
application for the plant-specific variations from TSTF-449.
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:
Exelon Generation Company, LLC, (EGC) has reviewed the proposed
no significant hazards consideration determination published on
March 2, 2005 (i.e., 70 FR 10298) as part of the consolidated line
item improvement process (CLIIP) item. EGC has concluded that the
proposed determination presented in the notice is applicable to
Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2,
and the determination is hereby incorporated by reference to satisfy
the requirements of 10 CFR 50.91 (a), except as discussed below.
The proposed amendment also revises the Technical Specification
Task Force (TSTF) Standard Technical Specification Change Traveler,
TSTF-449, ``Steam Generator Tube Integrity,'' Revision 4, version of
TS 5.5.9, Steam Generator Program, to exclude the portion of the
tube below 17 inches from the top of the hot leg tubesheet in the
Braidwood Station, Unit 2, and Byron Station, Unit 2, steam
generators from TS 5.5.9.d, ``Provisions for SG tube inspections.''
This proposed
[[Page 29677]]
license amendment request, in effect, redefines the Braidwood
Station, Unit 2, and Byron Station, Unit 2, primary pressure
boundary from the hot leg tube end weld to 17 inches below the top
of the hot leg tube sheet. This proposed license amendment also
deletes the current TS 5.5.9.e.6 and TS 5.5.9.e.10 allowance to use
Westinghouse laser welded sleeves as a SG tube repair method.
EGC has evaluated whether or not a significant hazards
consideration is involved with the proposed TS change by focusing on
the three criteria set forth in 10 CFR 50.92 as discussed below:
Criterion 1.--Does the proposed change involve a significant increase
in the probability or consequences of an accident previously evaluated?
Response: No.
The previously analyzed accidents are initiated by the failure
of plant structures, systems, or components. The proposed changes
that alter the SG inspection criteria and delete the allowance to
repair SG tubes using Westinghouse laser welded sleeves do not have
a detrimental impact on the integrity of any plant structure,
system, or component that initiates an analyzed event. The proposed
changes will not alter the operation of, or otherwise increase the
failure probability of any plant equipment that initiates an
analyzed accident. Therefore, the proposed change does not involve a
significant increase in the probability of an accident previously
evaluated.
Of the applicable accidents previously evaluated, the limiting
transients with consideration to the proposed changes to the SG tube
inspection criteria, are the SG tube rupture (SGTR) event and the
steam line break (SLB) accident.
During the SGTR event, the required structural integrity margins
of the SG tubes will be maintained by the presence of the SG
tubesheet. SG tubes are hydraulically expanded in the tubesheet
area. Tube rupture in tubes with cracks in the tubesheet is
precluded by the constraint provided by the tubesheet. This
constraint results from the hydraulic expansion process, thermal
expansion mismatch between the tube and tubesheet and from the
differential pressure between the primary and secondary side. Based
on this design, the structural margins against burst, discussed in
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR
[Pressurized Water Reactor] SG Tubes,'' are maintained for both
normal and postulated accident conditions.
The proposed changes do not affect other systems, structures,
components or operational features. Therefore, the proposed changes
result in no significant increase in the probability of the
occurrence of a SGTR accident.
At normal operating pressures, leakage from primary water stress
corrosion cracking (PWSCC) below the proposed limited inspection
depth is limited by both the tube-to-tubesheet crevice and the
limited crack opening permitted by the tubesheet constraint.
Consequently, negligible normal operating leakage is expected from
cracks within the tubesheet region. The consequences of an SGTR
event are affected by the primary-to-secondary leakage flow during
the event. Primary-to-secondary leakage flow through a postulated
broken tube is not affected by the proposed change since the
tubesheet enhances the tube integrity in the region of the hydraulic
expansion by precluding tube deformation beyond its initial
hydraulically expanded outside diameter.
The probability of a SLB is unaffected by the potential failure
of a SG tube as this failure is not an initiator for a SLB.
The consequences of a SLB are also not significantly affected by
the proposed changes. During a SLB accident, the reduction in
pressure above the tubesheet on the shell side of the SG creates an
axially uniformly distributed load on the tubesheet due to the
reactor coolant system pressure on the underside of the tubesheet.
The resulting bending action constrains the tubes in the tubesheet
thereby restricting primary-to-secondary leakage below the midplane.
Primary-to-secondary leakage from tube degradation in the
tubesheet area during the limiting accident (i.e., SLB) is limited
by flow restrictions resulting from the crack and tube-to-tubesheet
contact pressures that provide a restricted leakage path above the
indications and also limit the degree of potential crack face
opening as compared to free span indications. The primary-to-
secondary leak rate during postulated SLB accident conditions would
be expected to be less than that during normal operation for
indications near the bottom of the tubesheet (i.e., including
indications in the tube end welds). This conclusion is based on the
observation that while the driving pressure causing leakage
increases by approximately a factor of two, the flow resistance
associated with an increase in the tube-to-tubesheet contact
pressure, during a SLB, increases by up to approximately a factor of
three. While such a leakage decrease is logically expected, the
postulated accident leak rate could be conservatively bounded by
twice the normal operating leak rate if the increase in contact
pressure is ignored. Since normal operating leakage is limited to
less than 0.104 gpm [gallons per minute] (150 gpd [gallons per day])
per TS 3.4.13, ``RCS Operational Leakage,'' the associated accident
condition leak rate, assuming all leakage to be from lower tubesheet
indications, would be bounded by approximately 0.2 gpm. This value
is well within the assumed accident leakage rate of 0.5 gpm
discussed in Updated Final Safety Analysis Table 15.1-3,
``Parameters Used in Steam Line Break Analyses.'' Hence it is
reasonable to omit any consideration of inspection of the tube, tube
end weld, bulges/overexpansions or other anomalies below 17 inches
from the top of the hot leg tubesheet. Therefore, the consequences
of a SLB accident remain unaffected.
Based on the above discussion, the proposed changes do not
involve an increase in the consequences of an accident previously
evaluated.
Criterion 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 the use or installation of
new equipment and the currently installed equipment will not be
operated in a new or different manner. No new or different system
interactions are created and no new processes are introduced. The
proposed changes will not introduce any new failure mechanisms,
malfunctions, or accident initiators not already considered in the
design and licensing bases.
Based on this evaluation, the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
Criterion 3.--Does the proposed change involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes maintain the required structural margins of
the SG tubes for both normal and accident conditions. Nuclear Energy
Institute (NEI) 97-06, ``Steam Generator Program Guidelines,''
Revision 1 and Regulatory Guide (RG) 1.121, ``Bases for Plugging
Degraded PWR Steam Generator Tubes,'' are used as the bases in the
development of the limited hot leg tubesheet inspection depth
methodology for determining that SG tube integrity considerations
are maintained within acceptable limits. RG 1.121 describes a method
acceptable to the NRC for meeting General Design Criteria (GDC) 14,
``Reactor coolant pressure boundary,'' GDC 15, ``Reactor coolant
system design,'' GDC 31, ``Fracture prevention of reactor coolant
pressure boundary,'' and GDC 32, ``Inspection of reactor coolant
pressure boundary,'' by reducing the probability and consequences of
a SGTR. RG 1.121 concludes that by determining the limiting safe
conditions for tube wall degradation the probability and
consequences of a SGTR are reduced. This RG uses safety factors on
loads for tube burst that are consistent with the requirements of
Section III of the American Society of Mechanical Engineers (ASME)
Code.
For axially oriented cracking located within the tubesheet, tube
burst is precluded due to the presence of the tubesheet. For
circumferentially oriented cracking, Westinghouse letter LTR-CDME-
05-32, ``Limited Inspection of the Steam Generator Tube Portion
Within the Tubesheet at Byron Unit 2 and Braidwood Unit 2,''
Revision 2, dated August 2005, defines a length of degradation free
expanded tubing that provides the necessary resistance to tube
pullout due to the pressure induced forces, with applicable safety
factors applied. Application of the limited hot leg tubesheet
inspection depth criteria will preclude unacceptable primary-to-
secondary leakage during all plant conditions. The methodology for
determining leakage provides for large margins between calculated
and actual leakage values in the proposed limited hot leg tubesheet
inspection depth criteria.
Therefore, the proposed changes do not involve a significant
hazards consideration under the criteria set forth in 10 CFR
50.92(c).
The NRC staff has reviewed the licensee's analysis and, based on
this
[[Page 29678]]
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Brad J. Fewell, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett
Square, PA 19348.
NRC Branch Chief: Daniel S. Collins.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: January 25, 2006.
Description of amendment request: The proposed amendment would
revise the Updated Final Safety Analysis Report (UFSAR) to allow the
use of automatic load tap changers (LTCs) to operate in automatic mode
on the reserve auxiliary transformers (RATs) to compensate for
potential offsite power voltage fluctuations, in order to ensure that
acceptable voltage is maintained for safety related equipment.
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 requested change allows the automatic operation mode of the
LTC. The only accident previously evaluated for which the
probability is potentially affected by the change is the loss of
offsite power (LOOP). A failure of the LTC while in automatic
operation mode that results in decreased voltage to the ESS
[essential service system] buses could cause a LOOP. This could
occur in two ways. A failure of the LTC controller that results in
rapidly decreasing the voltage to the emergency buses is the most
severe failure mode. However, a backup controller is provided with
the LTC that makes this failure unlikely. A failure of the LTC
controller to respond to decreasing grid voltage is less severe,
since grid voltage changes occur slowly. In both of the above
potential failure modes, operators will take manual control of the
LTC to mitigate the effects of the failure. Thus, the probability of
a LOOP is not significantly increased.
The proposed change has no effect on the consequences of a LOOP,
since the emergency diesel generators provide power to safety
related equipment following a LOOP. The emergency diesel generators
are not affected by the proposed change.
The probability of other accidents previously evaluated is not
affected, since the proposed change does not affect the way plant
equipment is operated and thus does not contribute to the initiation
of any of the previously evaluated accidents.
The LTC is equipped with a backup controller, which controls the
LTC in the event of primary controller failure. Additionally,
operator action is available to prevent a sustained high voltage
condition from occurring. Damage due to over-voltage is time-
dependent. Therefore, damage of safety related equipment is
extremely unlikely, and the consequences of these accidents are not
significantly increased. The only way in which the consequences of
other previously evaluated accidents could be affected is if a
failure of the LTC, while in automatic operation mode, led to a
sustained high voltage condition, which resulted in damage to safety
related equipment that is used to mitigate an accident.
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 change involves functions that provide offsite
power to safety related equipment for accident mitigation. Thus, the
proposed change potentially affects the consequences of previously
evaluated accidents (as addressed in Question 1), but does not
result in any new mechanisms that could initiate damage to the
reactor and its principal safety barriers (i.e., fuel cladding,
reactor coolant system, or primary containment).
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 proposed change does not affect the inputs or assumptions of
any of the analyses that demonstrate the integrity of the fuel
cladding, reactor coolant system, or containment during accident
conditions. The allowable values for the degraded voltage protection
function are unchanged and will continue to ensure that the degraded
voltage protection function actuates when required, but does not
actuate prematurely to cause a LOOP. Automatic operation of the LTC
increases margin by reducing the potential for transferring to the
EDGs [emergency diesel generators] during an event.
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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelong Way, Kennett
Square, PA 19348.
NRC Branch Chief: Daniel S. Collins.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: February 10, 2006.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) 3.3.5.1, ``Emergency Core Cooling
System (ECCS) Instrumentation,'' to correct a Perry Nuclear Power Plant
(PNPP)-specific issue and establish consistency with the improved
standard technical specifications (ISTS). Specifically, Sub-actions
B.1.2.1 and B.1.2.2, which were added into PNPP TS 3.3.5.1 during the
ISTS conversion process, will be deleted. PNPP Required Action B.1 will
then match the ISTS Required Action B.1. As a result, actions with a 1-
hour completion time will only be required for the annulus exhaust gas
treatment (AEGT) system if a loss of initiation capability in both
divisions actually exists for an AEGT initiation function, as
originally intended.
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 Required Action within
the Technical Specifications. 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 does not
affect how the AEGT system is controlled, operated, or tested. The
intent of Required Action B.1 for the ECCS Instrumentation,
specifically, a loss of initiation capability check, is maintained
by the changes being proposed. The wording of Required Action B.1
ensures appropriate actions are taken when a loss of initiation
capability exists, by declaring the supported systems inoperable.
This action is consistent with the current requirements.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There are no physical modifications being made to any plant
system or component, and
[[Page 29679]]
the proposed change introduces no new method of operation for 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 ECCS Instrumentation Required Action continues
to ensure that a check is performed to determine if one or more of
the ECCS Instrumentation Functions has lost its capability to
actuate the Division 1 and 2 low-pressure ECCS, the AEGT subsystems,
and the associated diesel generators. It continues to direct
appropriate actions if such a loss of initiation capability is
found. Therefore, the necessary function of the Technical
Specification requirements is maintained, and 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: David W. Jenkins, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Daniel S. Collins.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: February 16, 2006.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) requirements related to steam
generator (SG) tube integrity. The change is consistent with NRC-
approved Revision 4 to Technical Specification Task Force (TSTF)
Standard Technical Specification Change Traveler, TSTF-449, ``Steam
Generator Tube Integrity.'' The availability of this TS improvement was
announced in the Federal Register on May 6, 2005 (70 FR 24126) as part
of the consolidated line item improvement process (CLIIP).
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on March 2,
2005 (70 FR 10298) as part of the CLIIP. The licensee affirmed the
applicability of the model NSHC determination in its application dated
February 16, 2006.
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:
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 [steam generator tube rupture] 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 [main steamline
break], 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 [Nuclear Energy Institute] 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 I-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 I-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 [500 gallons per
day or 720 gallons per day] in any one SG, and that the reactor
coolant activity levels of DOSE EQUIVALENT I-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
[[Page 29680]]
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: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Branch Chief: L. Raghavan.
Tennessee Valley Authority, Docket No. 50-259 , Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of amendment request: July 9, 2004 (TS-436).
Description of amendment request: The proposed amendment would
revise Technical Specification Surveillance Requirement 3.6.1.3.10 to
increase the allowed main steam isolation valve (MSIV) leak rate from
11.5 standard cubic feet per hour (scfh) per valve, to 100 scfh for
individual MSIVs with a 150 scfh combined leakage for all four main
steam lines.
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.
TVA proposes to utilize the main steam drain lines to
preferentially direct MSIV leakage to the main condenser. This drain
path takes advantage of the large volume of the steam lines and
condenser to provide holdup and plate-out of fission products that
may leak through the closed MSIVs. In this approach, the main steam
lines, steam drain piping, and the main condenser are used to
mitigate the consequences of an accident to limit potential doses
below the limits prescribed in 10 CFR 50.67(b)(2)(i) for the
exclusion area, 10 CFR 50.67(b)(2)(ii) for the low population zone,
and in 10 CFR 50.67(b)(2)(iii) for control room personnel.
Seismic verification walkdowns and evaluations of bounding
piping/supports were performed to demonstrate the main steam line
piping and components that comprise the Alternate Leakage Treatment
(ALT) path were rugged and able to perform the safety function of
MSIV leakage control following a Design Basis Earthquake (DBE).
Thus, it has been concluded the components in the MSIV alternate
leakage treatment flow path can be relied upon to maintain
structural integrity.
Therefore, the proposed amendment does not involve changes to
structures, components, or systems which would affect the
probability of an accident previously evaluated in the Browns Ferry
Updated Final Safety Analysis Report (UFSAR).
A plant-specific radiological analysis has been performed to
assess the effects of the proposed increase in MSIV leakage
acceptance criteria in terms of off-site doses and main control room
dose. The analysis shows the dose contribution from the proposed
increase in leakage acceptance criteria is acceptable compared to
doses limits prescribed in 10 CFR 50.67(b)(2)(i) for the exclusion
area, 10 CFR 50.67(b)(2)(ii) for the low population zone, and in 10
CFR 50.67(b)(2)(iii) for control room personnel.
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 require the use of the main steam piping
and the condenser to process MSIV leakage. This additional function
does not compromise the reliability of these systems. They will
continue to function as intended and not be subject to a failure of
a different kind than previously considered. In addition, MSIV
functionality will not be adversely impacted by the increased
leakage limit. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change to Surveillance Requirement 3.6.1.3.10, to
increase the allowable MSIV leakage, does not involve a significant
reduction in the margin of safety. The allowable leak rate specified
for the MSIVs is used to quantify a maximum amount of leakage
assumed to bypass containment. The results of the re-analysis
supporting these changes were evaluated against the dose limits
contained in 10 CFR 50.67(b)(2)(i) for the exclusion area, 10 CFR
50.67(b)(2)(ii) for the low population zone, and in 10 CFR
50.67(b)(2)(iii) for control room personnel. Sufficient margin
relative to the regulatory limits is maintained even when
conservative assumptions and methods are utilized. 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.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama
Date of amendment request: August 16, 2004 (TS--447).
Description of amendment request: The proposed amendment would
extend the channel calibration frequency requirements for
instrumentation in the high pressure coolant injection, reactor core
isolation cooling, and reactor water core isolation cooling systems.
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 extend the channel calibration surveillance
frequency of instrumentation used for the high area temperature
isolation of the high pressure coolant injection (HPCI), reactor
core isolation cooling (RCIC), and the reactor water clean-up (RWCU)
systems. The allowable trip point value for three sets of RCIC
instruments on each unit and for two sets of RWCU instruments on
Unit 1 are also revised. The calibration surveillance frequency is
extended to 24 months from 92 days (for the HPCI and RCIC high area
temperature instrumentation) and from 122 days (for the RWCU high
area temperature instrumentation). Under certain circumstances,
Technical Specifications (TS) SR [Surveillance Requirement] 3.0.2
would allow a maximum surveillance interval of 30 months for an SR
having a nominal 24-month performance frequency. Instrumentation
scaling and setpoint calculations performed in accordance with the
guidelines of Generic Letter 91-04 have shown that the reliability
of the affected protection instrumentation will be preserved for the
maximum allowable calibration surveillance interval. The Unit 1
instrumentation will be physically modified to be essentially
identical to that installed on Unit 2 and Unit 3 prior to restart of
Unit 1. Therefore, the proposed change does not involve a
significant increase in the
[[Page 29681]]
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 extend the channel calibration surveillance
frequency of instrumentation used for the high area temperature
isolation of the high pressure coolant injection (HPCI), reactor
core isolation cooling (RCIC), and the reactor water clean-up (RWCU)
systems. The allowable trip point value for three sets of RCIC
instruments on each unit and for two sets of RWCU instruments on
Unit 1 are also revised. The instrumentation will function in the
same way following the amendment as it functions currently. Hence,
the changes do not create the possibility of any new failure
mechanisms. Note that the Unit 1 instrumentation will be modified to
be essentially identical to that installed on Unit 2 and Unit 3
prior to restart of Unit 1. 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 proposed changes extend the channel calibration surveillance
frequency of instrumentation used for the high area temperature
isolation of the high pressure coolant injection (HPCI), reactor
core isolation cooling (RCIC), and the reactor water clean-up (RWCU)
systems. The allowable trip point value for three sets of RCIC
instruments on each unit and for two sets of RWCU instruments on
Unit 1 are also revised. Instrumentation scaling and setpoint
calculations performed in accordance with the guidelines of Generic
Letter 91-04 have shown safety margins are preserved with the
extended surveillance frequency and the revised TS allowable values.
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.
Tennessee Valley Authority (TVA), Docket No. 50-390, Watts Bar Nuclear
Plant (WBN), Unit 1, Rhea County, Tennessee
Date of amendment request: May 8, 2006 (TS-06-09).
Description of amendment request: The proposed amendment would
revise the limiting condition for operation for Technical Specification
(TS) Section 3.7.9, ``Ultimate Heat Sink.'' The maximum essential raw
cooling water (ERCW) temperature limit associated with Surveillance
Requirement 3.7.9.1 would increase from 85 degrees Fahrenheit ([deg]F)
to 88 [deg]F. This proposed change is based on evaluations of the ERCW
system and the ultimate heat sink (UHS) functions and maximum
temperatures that will satisfy the associated safety functions.
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 change to increase the UHS maximum temperature will
not adversely alter the function, design, or operating practices for
plant systems or components. The UHS is utilized to remove heat
loads from plant systems during normal and accident conditions. This
function is not expected or postulated to result in the generation
of any accident and continues to adequately satisfy the associated
safety functions with the proposed changes. Therefore, the
probability of an accident presently evaluated in the safety
analyses will not be increased. The heat loads, that the UHS is
designed to accommodate, have been evaluated with the higher
temperature limit. The result of these evaluations is that there is
existing margin associated with the systems that utilize the UHS for
normal and accident conditions. These margins are sufficient to
accommodate the postulated normal and accident heat loads with the
proposed changes to the UHS. Since the safety functions of the UHS
are maintained, the systems that ensure acceptable offsite dose
consequences will continue to operate as designed. The change in the
maximum calculated containment pressure associated with the design
basis loss-of-coolant-accident (LOCA) remains below the American
Society of Mechanical Engineers (ASME) Code design internal
pressure. Therefore, the consequence of any accident will be the
same as those previously analyzed.
Since the UHS safety function will continue to meet accident
mitigation requirements and limit dose consequences to acceptable
levels, TVA has concluded that the proposed TS 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 UHS function provides accident mitigation capabilities and
serves as a heat sink for normal and upset plant conditions; the UHS
is not an initiator of any accident. By allowing the proposed change
in the UHS temperature requirements, only the parameters for UHS
operation are changed while the safety functions of the UHS and
systems that transfer the heat sink capability continue to be
maintained. The proposed change does not impact the response of the
systems and components assumed in the safety analysis. Therefore,
the proposed change does not create the possibility of a new or
different kind of accident from any accident evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change has been evaluated for systems that are
needed to support accident mitigation functions as well as normal
operational evolutions. Operational margins were found to exist in
the systems that utilize the UHS capabilities such that these
proposed changes will not result in the loss of any safety function
necessary for normal or accident conditions. The ERCW system has
excess flow capacity that will accommodate the increased flows
necessary for the proposed temperature increase. While operating
margins have been reduced by the proposed changes, safety margins
have been maintained as assumed in the accident analyses for
postulated events. The proposed change results in an increase in the
maximum calculated containment peak pressure. However, the change in
the maximum calculated containment peak pressure associated with the
design basis LOCA is a small percentage of the margin between the
current maximum calculated containment peak pressure and the ASME
Code design internal pressure. This aspect of the proposed change
does not involve a significant reduction in a margin of safety.
Additionally, the proposed changes do not require any further
modification of component setpoints or operating provisions that are
necessary to maintain margins of safety established by the WBN
design (the shutdown board room chillers were physically modified to
operate properly at the 88 degree F UHS temperature). 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.
[[Page 29682]]
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: April 14, 2005, as supplemented by
letter dated December 21, 2005.
Description of amendment request: The amendment would revise the
Technical Specifications (TSs) by (1) adding a new TS 3.1.9, ``RCS
[Reactor Coolant System] Boron Limitations <500 [deg]F,'' and (2)
revising TS 3.3.1, ``Reactor Trip System (RTS) Instrumentation.''
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.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since there are
no design changes. The design of the reactor trip system (RTS)
instrumentation and engineered safety feature actuation system
(ESFAS) instrumentation will be unaffected and these protection
systems will continue to function in a manner consistent with the
plant design basis. All design, material, and construction standards
that were applicable prior to this amendment request will be
maintained.
The proposed changes will not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, and configuration of the facility or the manner in which
the plant is operated and maintained other than extending the
OPERABILITY requirements for RTS trip Function 2.b (Power Range
Neutron Flux--Low) to the upper portion of MODE 3. The proposed
changes will not alter or prevent the ability of structures,
systems, and components (SSCs) from performing their intended
functions to mitigate the consequences of an initiating event within
the assumed acceptance limits.
As discussed previously [in the application,] the proposed
change[s] will add more restrictive requirements in the form of a
new LCO [limiting condition for operation] 3.1.9 and an expanded LCO
Applicability for RTS trip Function 2.b, Power Range Neutron Flux--
Low, to provide mitigative capability in the event of an
uncontrolled RCCA [rod cluster control assembly] bank withdrawal
event postulated to occur during low power or subcritical (startup)
conditions.
There will be no change[s] to normal plant operating parameters
or accident mitigation performance. None of the proposed changes
will initiate any accidents; therefore, the probability of an
accident will not be increased. There will be no degradation in the
performance of, nor an increase in the number of challenges imposed
on, safety-related equipment assumed to function during an accident
situation.
All accident analysis acceptance criteria will continue to be
met with the proposed changes. The proposed changes will not affect
the source term, containment isolation, or radiological release
assumptions used in evaluating the radiological consequences of an
accident previously evaluated. The proposed changes will not alter
any assumptions or change any mitigation actions in the radiological
consequence evaluations in the FSAR [Final Safety Analysis Report
for Callaway]. The applicable radiological dose acceptance criteria
will continue to be met.
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.
There are no proposed design changes nor are there any changes
in the method by which any safety-related plant SSC performs its
safety function. [These changes] will not affect the normal method
of plant operation or change any operating parameters. No equipment
performance requirements will be affected other than the more
restrictive Applicability requirements being imposed on RTS trip
Function 2.b, Power Range Neutron Flux--Low, in the upper portion of
MODE 3. The proposed changes will not alter any assumptions made in
the safety analyses.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures will be introduced as a
result of this amendment. There will be no adverse effect or
challenges imposed on any safety-related system as a result of this
amendment.
The proposed amendment will not alter the design or performance
of the 7300 Process Protection System, Nuclear Instrumentation
System, or Solid State Protection System used in the plant
protection systems.
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.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions. There will be no impact on the overpower
limit, departure from nucleate boiling ratio (DNBR) limits, heat
flux hot channel factor (FQ), nuclear enthalpy rise hot
channel factor (F[Delta]H), loss of coolant accident peak cladding
temperature (LOCA PCT), peak local power density, or any other
margin of safety. The applicable radiological dose consequence
acceptance criteria will continue to be met.
The proposed changes do not eliminate any RTS or ESFAS
surveillances or alter the Frequency of surveillances required by
the Technical Specifications. More restrictive changes are proposed
by virtue of a new LCO 3.1.9 on [RCS] boron requirements when the
RCS temperature is below 500 [deg]F and by virtue of extending the
Applicability of RTS trip Function 2.b, Power Range Neutron Flux--
Low, to the upper portion of MODE 3. The nominal RTS and ESFAS trip
setpoints will remain unchanged. None of the acceptance criteria for
any accident analysis will be changed.
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.
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: April 20, 2006.
Brief description of amendment request: The proposed amendments
would reinstate the previous reactor coolant system pressure and
temperature limits, low temperature overpressure protection system
(LTOPS) setpoint, and (LTOPS) enable temperature basis that were
approved by the NRC staff on December 28, 1995, as License Amendments
Nos. 207 and 207 for Surry 1 and 2.
Date of publication of individual notice in Federal Register: April
28, 2006 (71 FR 25249)
Expiration date of individual notice: 30 day expiration date, May
30, 2006,
[[Page 29683]]
and 60 day expiration date, June 27, 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 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].
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: June 20, 2005.
Brief Description of amendments: The amendments revise the
Technical Specification (TS) Surveillance Requirement 3.6.1.6.2 of
3.6.1.6, ``Suppression Chamber-to-Drywell Vacuum Breakers'' for the
frequency of functionally testing the suppression chamber-to-drywell
vacuum breakers.
Date of issuance: May 5, 2006.
Effective date: May 5, 2006.
Amendment Nos.: 240 and 268.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the TS.
Date of initial notice in Federal Register: August 16, 2005 (70 FR
48202).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 5, 2006.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: August 18, 2005, as supplemented
by letter dated February 15, 2006.
Brief description of amendment: This amendment authorizes the use
of fire-resistive electrical cables in lieu of the alternatives
specified in Section C5.b.2 of Branch Technical Position Chemical
Engineering Branch 9.5-1 (NUREG-0800), `` Guidelines for Fire
Protection for Nuclear Power Plants,'' dated July 1981, for Fire Areas
12-A-CR, 1-A-CSRA, 1-A-CSRB, 1-A-SWGRA, 1-A-SWGRB, and 1-A-BAL-B.
Date of issuance: May 1, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No. 123.
Facility Operating License No. NPF-63: Amendment revises the
License.
Date of initial notice in Federal Register: November 8, 2005.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 1, 2006.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: July 14, 2005, as supplemented
January 11, 2006.
Brief description of amendment: The proposed change modifies the
Millstone Power Station, Unit No. 2 reactor coolant system heatup and
cooldown limits Technical Specification (TS) 3.4.9.1, ``Reactor Coolant
System''. The associated TS bases will be updated to address the
proposed change.
Date of issuance: May 3, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 292.
Facility Operating License No. DPR-65: The amendment revised the
TSs.
Date of initial notice in Federal Register: August 30, 2005 (70 FR
51379). The supplement dated January 11, 2006, provided clarifying
information that did not change the scope of the proposed amendment as
described in the original 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 May 3, 2006.
No significant hazards consideration comments received: No.
Duke Power Company, LLC Docket No. 50-287, Oconee Nuclear Station, Unit
3, Oconee County, South Carolina
Date of application of amendment: August 18, 2005, supplemented
September 15, 2005, and January 5 and April 6, 2006.
Brief description of amendment: The amendment revised Technical
Specifications 3.5.2.6 and 3.5.3.6 to accommodate the replacement of
the reactor building emergency sump suction inlet trash racks and
screens with strainers. Similar amendments were issued for Units 1 and
2 on November 1, 2005; however, the amendment for Unit 3 was not issued
at that time since the licensee had not completed its evaluation of the
impact of pipe whip, jet impingement and internally generated missiles
for Unit 3.
Date of Issuance: May 4, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 350.
Renewed Facility Operating License No. DPR-55: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: August 31, 2005 (70 FR
51852).
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 initial Federal
[[Page 29684]]
Register notice. The Commission's related evaluation of the amendment
is contained in a Safety Evaluation dated May 4, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of application for amendments: June 24, 2004.
Brief description of amendments: These amendments implement 25
generic Technical Specification (TS) changes previously approved by the
NRC staff as part of the Technical Specifications Task Force (TSTF).
The TSTF change travelers and proposed changes are:
1. TSTF-5, an administrative change to TS 2.2 to remove reporting
requirements that are already in the regulations 10 CFR, Sections 50.36
and 50.73;
2. TSTF-208, an extension of the time allowed to reach MODE 2 once
a TS 3.0.3 condition is identified, from the current 7 hours to 10
hours;
3. TSTFs-222 and 229, changes to TS 3.1.4 to allow scram time
testing on only affected rods when an outage is short and only a
limited number of fuel assemblies are moved and to require the Minimum
Critical Power Ratio to be determined after scram time testing;
4. TSTFs-297 and 227, changes to TSs 3.3.2.2, 3.3.4.1, and 3.3.4.2
to allow reactor feedwater pumps and main turbine valves to be removed
from service if their trip function is compromised;
5. TSTF-295, a clarification in Table 3.3.3.1-1 that penetration
flow paths, not just valve positions, are to be considered;
6. TSTF-275, a clarification Table 3.3.5.1-1 that certain emergency
core cooling system (ECCS) instrumentation needs to be operable when
ECCS and ECCS support systems are required to be operable;
7. TSTF-306, changes to TS 3.3.6.1 to allow penetration flow paths
to be opened intermittently under administrative controls and to set
apart the Traversing In-core Probe system isolation as a separate
function;
8. TSTF-416, changes to TSs 3.5.1 and 3.5.2 to allow the low
pressure coolant injection subsystems to be considered operable during
alignment and operation in the decay heat removal mode;
9. TSTF-17, a change to TS 3.6.1.2 to extend the containment air
lock interlock mechanism testing frequency from 6 months to 2 years to
coincide with refueling outage frequency;
10. TSTFs-30, 323, 45, 46, and 269, changes to TSs 3.6.1.3 and
3.6.4.2 related to primary and secondary containment isolation valve
completion times, isolation times, and status verification;
11. TSTF-322, a clarification in TS 3.6.4.1 of the intent of
secondary containment drawdown tests;
12. TSTF-276, Revision 2, a change to TS 3.8.1 to allow certain
emergency diesel generator (EDG) testing to continue even if the stated
power factor cannot be attained;
13. TSTF-404, a change to TS 3.1.8 to revise required actions when
one valve is inoperable in one or more scram discharge volume vent and
drain lines, as part of the consolidated line item improvement process;
14. TSTF-65 Revision 1, a change to allow the use of generic
organizational titles in the TSs, as opposed to plant-specific titles;
15. TSTF-299, a clarification in TS 5.2.2 of the intent of
refueling cycle intervals with respect to system leak test
requirements;
16. TSTF-279, a deletion in TS 5.5.6 of the reference to
``applicable supports'' as part of the description of the Inservice
Testing Program;
17. TSTF-118, a change to TS 5.5.9 to apply the provisions of
Surveillance Requirement (SR) 3.0.2 (25% extension interval) and SR
3.0.3 (missed surveillance actions) to EDG fuel oil testing
surveillances;
18. TSTF-106, Revision 1, a clarification in TS 5.5.9 that the
American Society for Testing and Materials standard for EDG fuel oil
applies only to new fuel being received; and
19. TSTF-152, a change to the Radioactive Effluent Release Report
to ensure that a common report for both units combines sections common
to both units.
Date of issuance: May 10, 2006.
Effective date: As of the date of issuance, to be implemented
within 90 days.
Amendments Nos.: 259 and 262.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The
amendments revised the TSs.
Date of initial notice in Federal Register: September 28, 2004 (69
FR 57985) and October 26, 2004 (69 FR 62476).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 10, 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: December 19, 2006.
Description of amendment request: The amendment deletes Technical
Specification (TS) 6.8.1.2a, ``Occupational Radiation Exposure Report
[ORER],'' TS 6.8.1.2.c, regarding challenges to pressurizer relief and
safety valves and TS 6.8.1.5, ``Monthly Operating Report [MOR],'' as
described in the Notice of Availability published in the Federal
Register on June 23, 2004 (69 FR 35067).
Date of issuance: May 5, 2006.
Effective date: As of its date of issuance, and shall be
implemented within 90 days.
Amendment No.: 109.
Facility Operating License No. NPF-86: The amendment revised the
TSs.
Date of initial notice in Federal Register: February 14, 2006 (71
FR 7808).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 5, 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: October 12, 2005.
Brief description of amendment: The amendment revised Technical
Specification (TS) Section 3.4.9, ``RCS [reactor coolant system]
Pressure and Temperature (P/T) Limits,'' curves 3.4.9-1, ``Pressure/
Temperature Limits for Non-Nuclear Heatup or Cooldown Following Nuclear
Shutdown,'' 3.4.9-2, ``Pressure/Temperature Limits for Inservice
Hydrostatic and Inservice Leakage Tests, and 3.4.9-3, ``Pressure/
Temperature Limits for Criticality,'' to remove the cycle operating
restriction and replace it with a limitation of 30 effective full-power
years (EFPY).
Date of issuance: April 27, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 219.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 3, 2006 (71 FR
150).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 27, 2006.
No significant hazards consideration comments received: No.
[[Page 29685]]
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: December 30, 2005.
Brief description of amendment: The amendment established a
combined leakage rate limit for the sum of the four main steam line
leakage rates that is equal to four times the current individual main
steam isolation valve leakage rate limit.
Date of issuance: May 2, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 220.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 28, 2006 (71
FR 10073)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 2, 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: January 30, 2006.
Brief description of amendment: The amendment allows a delay time
for entering a supported system Technical Specification (TS) when the
inoperability is due solely to an inoperable snubber, if risk is
assessed and managed consistent with the program in place for complying
with the requirements of 10 CFR 50.65(a)(4). Limiting Condition for
Operation (LCO) 3.0.8 is added to the TS to provide this allowance and
define the requirements and limitations for its use.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-372, Revision 4. The NRC staff
issued a notice of opportunity for comment in the Federal Register on
November 24, 2004 (69 FR 68412), on possible amendments concerning
TSTF-372, including a model safety evaluation and model no significant
hazards consideration (NSHC) determination, using the consolidated line
item improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on May 4, 2005 (70 FR 23252). The
licensee affirmed the applicability of the following NSHC determination
in its application dated January 30, 2006.
Date of issuance: May 2, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 221.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 28, 2006 (71
FR 10074).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 2, 2006.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant (MNGP), Wright County, Minnesota
Date of application for amendment: April 29, 2004, as supplemented
on November 23, 2004; January 20, February 28, April 12, 2005; and
March 10, 2006.
Brief description of amendment: The amendment revised the MNGP
licensing basis by selectively implementing the alternative source term
for the postulated fuel handling accident, leading to revision of
portions of the Technical Specifications to reflect this change in
licensing basis.
Date of issuance: April 24, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 145.
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 18, 2005 (70 FR
2891)
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 April 24, 2006.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: August 11, 2005.
Brief description of amendment: The change allows a delay time for
entering a supported system Technical Specification (TS) when the
inoperability is due solely to an inoperable snubber, if risk is
assessed and managed consistent with the program in place for complying
with the requirements of 10 CFR 50.65(a)(4). Limiting Condition for
Operation (LCO) 3.0.8 is added to the TS to provide this allowance and
define the requirements and limitations for its use.
Date of issuance: March 1, 2006.
Effective date: As of its date of issuance and shall be implemented
within 120 days of issuance.
Amendment No.: 238.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: December 6, 2005 (70 FR
72674)
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated March 1, 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.
Brief description of amendments: The change allows a delay time for
entering a supported system Technical Specification (TS) when the
inoperability is due solely to an inoperable snubber, if risk is
assessed and managed consistent with the program in place for complying
with the requirements of 10 CFR 50.65(a)(4). Limiting Condition for
Operation (LCO) 3.0.8 is added to the TS to provide this allowance and
define the requirements and limitations for its use.
Date of issuance: March 7, 2006.
Effective date: As of its date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: Unit 1--185; Unit 2--187
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 Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 7, 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: December 9, 2004, as
supplemented by letters dated November 18 and December 5, 2005.
Brief description of amendment: The amendment authorizes
modification to
[[Page 29686]]
the Updated Final Safety Analysis Report (UFSAR) to include a revision
to the methodology for splicing reinforcing steel bars during
restoration of the Unit 1 concrete shield building dome as part of the
steam generator replacement project.
Date of issuance: April 27, 2006.
Effective date: As of the date of issuance and shall be implemented
as part of the next UFSAR update made in accordance with 10 CFR
50.71(e).
Amendment No. 60.
Facility Operating License No. NPF-90: Amendment authorizes
revision of the Updated Final Safety Analysis Report.
Date of initial notice in the Federal Register: January 4, 2005 (70
FR 405). 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 April 27, 2006.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: October 30, 2003, as supplemented by
letters dated August 31 and November 18, 2005, and March 6, 2006.
Brief description of amendment: The amendment increases the
completion times (CTs) for Technical Specification (TS) 3.8.1, ``AC
Sources--Operating,'' and adds requirements on the diesel generators at
the Sharpe Station when a diesel generator at Wolf Creek Generating
Station is in an extended CT greater than 72 hours. The proposed
changes to TS 3.8.9, ``Distribution Systems--Operating,'' are
withdrawn. The amendment also revises a page in the license and adds
conditions to Appendix D, ``Additional Conditions,'' of the license.
Date of issuance: April 26, 2006.
Effective date: As of its date of issuance and shall be implemented
within 90 days of the date of issuance.
Amendment No.: 163.
Facility Operating License No. NPF-42. The amendment revised the
license including Appendix D, ``Additional Conditions,'' and Appendix
A, ``Technical Specifications.''
Date of initial notice in Federal Register: January 6, 2004 (69 FR
700).
The supplemental letters dated August 31 and November 18, 2005, and
March 2, 2006, provided additional information that clarified the
application, 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 published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 26, 2006.
No significant hazards consideration comments received: No
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: November 3, 2005, and supplemental
letters dated February 21 and March 28, 2006.
Brief description of amendment: The amendment revised the Technical
Specifications associated with steam generator tube integrity
consistent with Revision 4 to Technical Specification Task Force (TSTF)
Standard Technical Specification Change Traveler, TSTF-449, ``Steam
Generator Tube Integrity.'' A notice of availability for this TS
improvement using the consolidated line item improvement process was
published in the Federal Register on May 6, 2005 (70 FR 24126).
Date of issuance: May 8, 2006.
Effective date: The license amendment is effective as of its date
of issuance and shall be implemented prior to the entry into Mode 5 in
the restart from Refueling Outage 15, which is scheduled to begin in
October 2006.
Amendment No.: 164.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 6, 2005 (70 FR
72676) The supplemental letters dated February 21 and March 28, 2006,
provided additional clarifying information, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 8, 2006.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 15th day of May 2006.
For the Nuclear Regulatory Commission
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 06-4736 Filed 5-22-06; 8:45 am]
BILLING CODE 7590-01-P