[Federal Register Volume 71, Number 137 (Tuesday, July 18, 2006)]
[Notices]
[Pages 40742-40759]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-6246]
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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
[[Page 40743]]
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 June 23, 2006 to July 6, 2006. The last
biweekly notice was published on July 5, 2006 (71 FR 38180).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
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
[[Page 40744]]
when the hearing is held. If the final determination is that the
amendment request involves no significant hazards consideration, the
Commission may issue the amendment and make it immediately effective,
notwithstanding the request for a hearing. Any hearing held would take
place after issuance of the amendment. If the final determination is
that the amendment request involves a significant hazards
consideration, any hearing held would take place before the issuance of
any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: May 15, 2006.
Description of amendment request: The amendment would revise the
Technical Specification (TS) requirements related to steam generator
tube integrity. The proposed changes are generally consistent with
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 proposed
amendment includes changes to licensing pages to delete License
Condition 2.c.(8), ``Repaired Steam Generators;'' changes to TS 3.1.6,
``LEAKAGE;'' changes to TS Section 3.1.1.2, ``Steam Generators and
Steam Generator (SG) Tube Integrity;'' revising TS Section 4.19,
``Steam Generator (SG) Tube Integrity;'' adding new TS 6.9.6, ``Steam
Generator Tube Inspection Report;'' and adding new TS 6.19, ``Steam
Generator (SG) Program.''
Basis for proposed no significant hazards consideration
determination (NSHC): The NRC staff published 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 NSHC determination, using the CLIIP. 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 following NSHC determination in its application dated May 15,
2006. As required by 10 CFR 50.91(a), an analysis of the issue of no
significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The proposed change requires a SG Program that includes
performance criteria that will provide reasonable assurance that the
SG tubing will retain integrity over the full range of operating
conditions (including startup, operation in the power range, hot
standby, cooldown and all anticipated transients included in the
design specification). The SG performance criteria are based on tube
structural integrity, accident induced leakage, and operational
LEAKAGE.
A Steam Generator Tube Rupture (SGTR) event is one of the
design-basis accidents that are analyzed as part of a plant's
licensing basis. In the analysis of a SGTR event, a bounding primary
to secondary LEAKAGE rate equal to the operational LEAKAGE rate
limits in the licensing basis plus the LEAKAGE rate associated with
a double-ended rupture of a single tube is assumed.
For other design-basis accidents such as Main Steam Line Break
(MSLB), rod ejection, and reactor coolant pump locked rotor the
tubes are assumed to retain their structural integrity (i.e., they
are assumed not to rupture). These analyses typically assume that
primary to secondary LEAKAGE for all SGs is 1 gallon per minute or
increases to 1 gallon per minute as a result of accident-induced
stresses. The accident-induced leakage criterion introduced by the
proposed changes accounts for tubes that may leak during design-
basis accidents. The accident-induced leakage criterion limits this
leakage to no more than the value assumed in the accident analysis.
The SG performance criteria proposed change to the TSs
identifies 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 TSs. 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 the primary-to-secondary leak rate after the
accident is 1 gallon per minute with no more than 500 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
[[Page 40745]]
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 change does not affect the
consequences of an MSLB, rod ejection, or a reactor coolant pump
locked rotor event, or other previously-evaluated accident.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident From Any Previously
Evaluated.
The proposed performance-based requirements are an improvement
over the requirements imposed by the current technical
specifications. Implementation of the proposed SG Program will not
introduce any adverse changes to the plant design-basis or
postulated accidents resulting from potential tube degradation. The
result of the implementation of the SG Program will be an
enhancement of SG tube performance. Primary to secondary LEAKAGE
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
The SG tubes in pressurized-water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the SG Program to manage SG tube
inspection, assessment, repair, and plugging. The requirements
established by the SG Program are consistent with those in the
applicable design codes and standards and are an improvement over
the requirements in the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TSs.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Mr. Brad Fewell, Assistant General Counsel,
Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA
19348.
NRC Branch Chief: Darrell J. Roberts.
Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: June 1, 2006.
Description of amendment request: The proposed amendments would
revise the Updated Final Safety Analysis Report (UFSAR) to incorporate
the use of a fiber-reinforced polymer (FRP) system to strengthen
existing masonry walls against tornado effects.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Response: Physical protection from a tornado event is a design
basis criterion rather than a requirement of a previously analyzed
UFSAR accident analysis.
The current licensing basis (CLB) for Oconee states that
systems, structures, and components (SSC's) required to shut down
and maintain the units in a shutdown condition will not fail as a
result of damage caused by natural phenomena.
The in-fill masonry walls to be strengthened using an FRP system
are passive, non-structural elements. The use of an FRP system on
existing Auxiliary Building masonry walls will allow them to resist
uniform pressure loads resulting from a tornado and will not
adversely affect the structure's ability to withstand other design
basis events such as earthquakes or fires. Therefore, the proposed
use of FRP on existing masonry walls will not significantly increase
the probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
Response: The final state of the FRP system is passive in nature
and will not initiate or cause an accident. More generally, this
understanding supports the conclusion that the potential for new or
different kinds of accidents is not created.
3. Involve a significant reduction in a margin of safety.
Response: The application of an FRP system to existing auxiliary
building masonry walls will either act to restore the margin of
safety described in the UFSAR, e.g., the Unit 3 Control Room north
wall, or enhance the margin of safety, e.g., the West Penetration
Room walls, by increasing the walls' ability to resist tornado-
induced differential pressure and/or tornado wind. Consequently,
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: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Branch Chief: Evangelos C. Marinos.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: May 22, 2006.
Description of amendment request: The proposed license amendment
request would revise: (1) Surveillance Requirement (SR) 3.8.1.11 to
remove the MODE restriction from Note 2 for Diesel Generator (DG)-3
only, (2) SR 3.8.1.12 to remove the MODE restriction from Note 2 for
DG-3 only, (3) SR 3.8.1.16 to remove the MODE restriction from the Note
for DG-3 only, and (4) Revise SR 3.8.1.19 to remove the MODE
restriction from Note 2 for DG-3 only.
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 DG and its associated emergency loads are accident
mitigating features, not accident initiating equipment. Therefore,
there will be no impact on any accident probabilities by the
approval of the requested amendment. The design of plant equipment
is not being modified by these proposed changes. The capability of
DG-1 and DG-2 to supply power to their safety related buses as
designed will not be compromised by permitting performance of DG-3
testing during power operations. Columbia's Technical Specifications
require the RCIC [reactor core isolation cooling] system to be
operable whenever this testing is performed at power. This ensures
that the high-pressure injection function is maintained during the
time the HPCS injection valve is disabled
[[Page 40746]]
during testing. In the event of a design basis accident during
testing, the HPCS [high-pressure core spray] system could be
returned to service well within the 14-day outage time allowed by
Technical Specifications. Additionally, the ability of the Standby
Liquid Coolant (SLC) system to perform its design safety function
would not be affected because SLC is connected downstream of the
HPCS injection valve. Therefore, there would be no significant
impact on any accident consequences.
Based on the above, the proposed change to permit certain DG
surveillance tests to be performed during plant operation will have
no effect on accident probabilities or consequences. Therefore, the
proposed change does 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.
No new accident causal mechanisms would be introduced as a
result of NRC approval of this amendment request since no changes
are being made to the plant that would introduce any new accident
causal mechanisms. Equipment will be operated in the same
configuration with the exception of the plant mode in which the
testing is conducted. This amendment request does not impact any
plant systems that are accident initiators; neither does it
adversely impact any accident mitigating systems.
Based on the above, implementation of the proposed changes would
not create the possibility of a new or different kind of accident
from any accident 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.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their design functions
during and following an accident situation. These barriers include
the fuel cladding, the reactor coolant system, and the containment
system. The proposed changes to the testing requirements for the DG
do not affect the operability requirements for the DG, as
verification of such operability will continue to be performed as
required. Continued verification of operability supports the
capability of the DG to perform its required function of providing
emergency power to plant equipment that supports or constitutes the
fission product barriers. Consequently, the performance of these
fission product barriers will not be impacted by implementation of
this proposed amendment. In addition, the proposed changes involve
no changes to setpoints or limits established or assumed by the
accident analysis. On this, and the above basis, no safety margins
will be impacted.
Energy Northwest concludes that there is no significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao.
Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant,
Unit No. 1, St. Lucie County, Florida
Date of amendment request: April 24, 2006.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) consistent with the NRC-
approved Revision 4 to TS Task Force (TSTF) Standard TS Change
Traveler, TSTF-449, ``Steam Generator Tube Integrity.''
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 following
NSHC determination in its application dated April 24, 2006. Basis for
proposed no significant hazards consideration determination: As
required by 10 CFR 50.91(a), an analysis of the issue of no significant
hazards consideration is presented below:
Criterion 1--The proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change requires a SG [Steam Generator] 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[n] 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[n] 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[s] to the TS[s]
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[s]. 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
[[Page 40747]]
kind of accident from any previously evaluated.
The proposed performance based requirements are an improvement
over the requirements imposed by the current technical
specifications. Implementation of the proposed SG Program will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
result of the implementation of the SG Program will be an
enhancement of SG tube performance. Primary to secondary LEAKAGE
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
Criterion 3--The proposed change does not involve a significant
reduction in the margin of safety.
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes. Steam generator
tube integrity is a function of the design, environment, and the
physical condition of the tube. The proposed change does not affect
tube design or operating environment. The proposed change is
expected to result in an improvement in the tube integrity by
implementing the SG Program to manage SG tube inspection,
assessment, repair, and plugging. The requirements established by
the SG Program are consistent with those in the applicable design
codes and standards and are an improvement over the requirements in
the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TS.
Based upon the reasoning presented above 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: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Michael L. Marshall, Jr.
Florida Power and Light Company, Docket No. 50-389, St. Lucie Plant,
Unit No. 2, St. Lucie County, Florida
Date of amendment request: May 25, 2006.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) consistent with the NRC-
approved Revision 4 to TS Task Force (TSTF) Standard TS Change
Traveler, TSTF-449, ``Steam Generator Tube Integrity.''
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 following
NSHC determination in its application dated May 25, 2006.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change requires a SG [Steam Generator] 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[n] 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[n] 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[s] to the TS[s]
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[s]. 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
[[Page 40748]]
enhancement of SG tube performance. Primary to secondary LEAKAGE
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
Criterion 3--The proposed change does not involve a significant
reduction in the margin of safety.
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes. Steam generator
tube integrity is a function of the design, environment, and the
physical condition of the tube. The proposed change does not affect
tube design or operating environment. The proposed change is
expected to result in an improvement in the tube integrity by
implementing the SG Program to manage SG tube inspection,
assessment, repair, and plugging. The requirements established by
the SG Program are consistent with those in the applicable design
codes and standards and are an improvement over the requirements in
the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TS.
Based upon the reasoning presented above 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: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Michael L. Marshall, Jr.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: April 27, 2006.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) consistent with the NRC-
approved Revision 4 to TS Task Force (TSTF) Standard TS Change
Traveler, TSTF-449, ``Steam Generator Tube Integrity.''
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 following
NSHC determination in its application dated April 27, 2006.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change requires a SG [Steam Generator] 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[n] 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[n] 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[s] to the TS[s]
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[s]. 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.
[[Page 40749]]
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
Criterion 3--The proposed change does not involve a significant
reduction in the margin of safety.
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes. Steam generator
tube integrity is a function of the design, environment, and the
physical condition of the tube. The proposed change does not affect
tube design or operating environment. The proposed change is
expected to result in an improvement in the tube integrity by
implementing the SG Program to manage SG tube inspection,
assessment, repair, and plugging. The requirements established by
the SG Program are consistent with those in the applicable design
codes and standards and are an improvement over the requirements in
the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TS.
Based upon the reasoning presented above 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: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Michael L. Marshall, Jr.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: January 18, 2006.
Description of amendment request: The proposed amendment would
delete the reference to the hydrogen monitors in Technical
Specification (TS) 3.6.11, ``Accident Monitoring Instrumentation''
consistent with the NRC-approved Industry/Technical Specification Task
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
447, ``Elimination of Hydrogen Recombiners and Change to Hydrogen and
Oxygen Monitors.''
The NRC staff issued a notice of availability of ``Model
Application Concerning Technical Specification Improvement To Eliminate
Hydrogen Recombiner Requirement, and Relax the Hydrogen and Oxygen
Monitor Requirements for Light Water Reactors Using the Consolidated
Line Item Improvement Process (CLIIP)'', in the Federal Register on
September 25, 2003 (68 FR 55416). The notice included a model safety
evaluation (SE), a model no significant hazards consideration (NSHC)
determination, and a model application.
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, by confirming the applicability of the model NSHC
determination to NMP-1 and incorporating it by reference in its
application. The model NSHC determination 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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen [and oxygen] monitors are no longer required to mitigate
design-basis accidents and, therefore, the hydrogen monitors do not
meet the definition of a safety-related component as defined in 10
CFR 50.2. RG [Regulatory Guide] 1.97 Category 1, is intended for key
variables that most directly indicate the accomplishment of a safety
function for design-basis accident events. The hydrogen [and oxygen]
monitors no longer meet the definition of Category 1 in RG 1.97. As
part of the rulemaking to revise 10 CFR 50.44 the Commission found
that Category 3, as defined in RG 1.97, is an appropriate
categorization for the hydrogen monitors because the monitors are
required to diagnose the course of beyond design-basis accidents.
[Also, as part of the rulemaking to revise 10 CFR 50.44, the
Commission found that Category 2, as defined in RG 1.97, is an
appropriate categorization for the oxygen monitors, because the
monitors are required to verify the status of the inert
containment.]
The regulatory requirements for the hydrogen [and oxygen]
monitors can be relaxed without degrading the plant emergency
response. The emergency response, in this sense, refers to the
methodologies used in ascertaining the condition of the reactor
core, mitigating the consequences of an accident, assessing and
projecting offsite releases of radioactivity, and establishing
protective action recommendations to be communicated to offsite
authorities. Classification of the hydrogen monitors as Category 3,
[classification of the oxygen monitors as Category 2] and removal of
the hydrogen [and oxygen] monitors from TS will not prevent an
accident management strategy through the use of the SAMGs [severe
accident management guidelines], the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen [and oxygen] monitor
requirements, including removal of these requirements from TS, does
not involve a significant increase in the probability or the
consequences of any accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident From Any Previously
Evaluated.
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen [and oxygen] monitor requirements,
including removal of these requirements from TS, will not result in
any failure mode not previously analyzed. The hydrogen recombiner
and hydrogen [and oxygen] monitor equipment was intended to mitigate
a design-basis hydrogen release. The hydrogen recombiner and
hydrogen [and oxygen] monitor equipment are not considered accident
precursors, nor does their existence or elimination have any adverse
impact on the pre-accident state of the reactor core or post
accident confinement of radionuclides within the containment
building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in [a] Margin of Safety.
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen [and oxygen] monitor requirements,
including removal of these requirements from TS, in light of
existing plant equipment, instrumentation, procedures, and programs
that provide effective mitigation of and recovery from reactor
accidents, results in a neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that
[[Page 40750]]
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI [Three Mile Island],
Unit 2 accident can be adequately met without reliance on safety-
related hydrogen monitors.
[Category 2 oxygen monitors are adequate to verify the status of
an inerted containment.]
Therefore, this change does not involve a significant reduction
in [a] margin of safety. [The intent of the requirements established
as a result of the TMI, Unit 2 accident can be adequately met
without reliance on safety-related oxygen monitors.] Removal of
hydrogen [and oxygen] monitoring from TS will not result in a
significant reduction in their functionality, reliability, and
availability.
The NRC staff has reviewed the model NSHC determination and its
applicability to NMP-1. Based on this review, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: Richard J. Laufer.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: June 6, 2006.
Description of amendment request: The proposed amendments would
revise the design basis as described in the Point Beach Nuclear Plant
Final Safety Analysis Report (FSAR) by incorporating an updated
analysis for satisfying the reactor vessel Charpy upper-shelf energy
requirements of 10 CFR part 50, Appendix G, Section IV.A.1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Would the proposed amendment involve a significant increase
in the probability or consequences of any accident previously
evaluated?
The proposed change incorporates the updated analysis for
satisfying the reactor vessel Charpy upper-shelf energy requirements
of 10 CFR part 50, Appendix G, Section IV.A.1 into the FSAR. The
proposed change does not adversely affect accident initiators or
precursors nor alter the design assumptions, conditions, or the
manner in which the plant is operated and maintained. The proposed
change does not alter or prevent the ability of structures, systems,
and components from performing their intended function to mitigate
the consequences of an initiating event within the assumed
acceptance limits. The proposed change does not affect the source
term, containment isolation, or radiological release assumptions
used in evaluating the radiological consequences of an accident
previously evaluated. Further, the proposed change does not increase
the types or amounts of radioactive effluent that may be released
offsite, nor significantly increase individual or cumulative
occupational/public radiation exposures. The proposed change is
consistent with safety analysis assumptions and resultant
consequences. Therefore, it is concluded that this change does not
significantly increase the probability of occurrence of an accident
previously evaluated.
2. Would the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
The proposed change incorporates the updated analysis for
satisfying the reactor vessel Charpy upper-shelf energy requirements
of 10 CFR part 50, Appendix G, Section IV.A.1 into the FSAR. The
change does not impose any new or different requirements or
eliminate any existing requirements. The change does not alter
assumptions made in the safety analysis. The proposed change is
consistent with the safety analysis assumptions and current plant
operating practice. Therefore, the proposed change would not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. Would the proposed amendment result in a significant
reduction in a margin of safety?
The proposed change incorporates the updated analysis for
satisfying the reactor vessel Charpy upper-shelf energy requirements
of 10 CFR part 50, Appendix G, Section IV.A.1 into the FSAR. The
proposed change does not alter the manner in which safety limits,
limiting safety system settings or limiting conditions for operation
are determined. The setpoints at which protective actions are
initiated are not altered by the proposed change. Therefore, the
proposed amendment does not result in 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: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Branch Chief: L. Raghavan.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: May 30, 2006.
Description of amendment request: The proposed amendment would
revise the Fort Calhoun Station, Unit 1 (FCS) Technical Specification
(TS) requirements related to steam generator 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).
Omaha Public Power District (OPPD) also proposes to change the FCS
TS by deleting the sleeving repair alternative to plugging for steam
generator tubes. The FCS replacement steam generators (RSGs) to be
installed during the fall of 2006 are manufactured by Mitsubishi Heavy
Industries, Ltd. (MHI). The change is being requested because OPPD has
determined that the sleeving repair alternative to plugging will not be
used for the MHI RSGs.
Basis for proposed no significant hazards consideration
determination: OPPD stated that it had reviewed the proposed no
significant hazards consideration determination published on March 2,
2005 (70 FR 10298), as part of the CLIIP. OPPD has concluded that the
proposed determination presented in the notice is applicable to FCS and
the determination is incorporated by reference to satisfy the
requirements of 10 CFR 50.91(a). 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 elimination from the TS surveillance requirements of leak
tight sleeves as a repair method alternative to plugging defective
steam generator tubes does not introduce an initiator to any
previously evaluated accident. The frequency or periodicity of
performance of the remaining surveillance requirements for steam
generator tubes (including plugged tubes) is not affected by this
change. Elimination of the tube repair method has no effect on the
consequences of any previously evaluated accident. The
[[Page 40751]]
proposed changes will not prevent safety systems from performing
their accident mitigation function as assumed in the safety
analysis.
Therefore, this change does not involve a significant increase
in the probability or consequences of any 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 only affects the TS surveillance
requirements. The proposed change is a result of installation of
RSGs. The proposed change will eliminate a steam generator tube
repair alternative which cannot be utilized or credited for the
RSGs. This change will not alter assumptions made in the safety
analysis and licensing bases and will not create new or different
systems interactions.
Therefore, this 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 deletes surveillance requirements for a
steam generator tube repair alternative which will no longer be
necessary or applicable. The remaining TS steam generator tube
surveillance requirements, including inspection and plugging
requirements, will continue to maintain the applicable margin of
safety.
Therefore, this TS change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn,
1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: David Terao
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: May 30, 2006.
Description of amendment requests: The proposed amendment would
revise the Technical Specifications (TSs) 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 includes changes to the TS
definition of Leakage, TS 3.4.13, ``RCS [Reactor Coolant System]
Operational Leakage,'' TS 5.5.9, ``Steam Generator (SG) Tube
Surveillance Program,'' TS 5.6.10, ``Steam Generator (SG) Tube
Inspection Report,'' and adds TS 3.4.17, ``Steam Generator (SG) Tube
Integrity.'' The proposed changes are necessary in order to implement
the guidance for the industry initiative on NEI 97-06, ``Steam
Generator Program Guidelines.''
The NRC staff issued a notice of 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 following
NSHC determination in its application dated May 30, 2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The proposed change requires an 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 steam generator tube rupture (SGTR) event is one of the
design-basis accidents that are analyzed as part of a plant's
licensing basis. In the analysis of an 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 a main steamline break
(MSLB), rod ejection, and reactor coolant pump locked rotor, the
tubes are assumed to retain their structural integrity (i.e., they
are assumed not to rupture). These analyses typically assume that
primary to secondary LEAKAGE for all SGs are 1 gallon per minute or
increases to 1 gallon per minute as a result of accident-induced
stresses. The accident-induced leakage criterion introduced by the
proposed changes accounts for tubes that may leak during design-
basis accidents. The accident-induced leakage criterion limits this
leakage to no more than the value assumed in the accident analysis.
The SG performance criteria proposed change to the TS identify
the standards against which tube integrity is to be measured.
Meeting the performance criteria provides reasonable assurance that
the SG tubing will remain capable of fulfilling its specific safety
function of maintaining reactor coolant pressure boundary integrity
throughout each operating cycle and in the unlikely event of a
design-basis accident. The performance criteria are only a part of
the SG Program required by the proposed change to the TS. The
program, defined by NEI 97-06, ``Steam Generator Program
Guidelines,'' includes a framework that incorporates a balance of
prevention, inspection, evaluation, repair, and leakage monitoring.
The proposed changes do not, therefore, significantly increase the
probability of an accident previously evaluated.
The consequences of design-basis accidents are, in part,
functions of the DOSE EQUIVALENT 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 an 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,
[[Page 40752]]
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 the primary system's pressure and inventory. As part of the
reactor coolant pressure boundary, the SG tubes are unique in that
they are also relied upon as a heat transfer surface between the
primary and secondary systems such that residual heat can be removed
from the primary system. In addition, the SG tubes isolate the
radioactive fission products in the primary coolant from the
secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the SG Program to manage SG tube
inspection, assessment, repair, and plugging. The requirements
established by the SG Program are consistent with those in the
applicable design codes and standards and are an improvement over
the requirements in the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TS.
The NRC staff proposes to determine that the amendment requests
involve no significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Branch Chief: David Terao.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: May 1, 2006.
Description of amendment request: The proposed amendment would
eliminate the requirement for a power range, neutron flux, high
negative rate trip and delete the references to this trip as functional
Unit 4 in Salem Generating Station (Salem) Unit Nos. 1 and 2 Technical
Specification (TS) Table 2.2-1, ``Reactor Trip System Instrumentation
Trip Setpoints,'' TS Table 3.3-1, ``Reactor Trip System
Instrumentation,'' TS Table 3.3-2, ``Reactor Trip System
Instrumentation Response Times,'' and TS Table 4.3-1, ``Reactor Trip
System Instrumentation Surveillance Requirements [SRs].'' The proposed
changes are consistent with the methodology presented in the
Westinghouse Topical Report WCAP-11394-P-A, ``Methodology for the
Analysis of the Dropped Rod Event,'' which has been reviewed by the NRC
and found acceptable for referencing in license applications. The
amendment also would involve the correction of errata in the TS for
Salem Unit Nos. 1 and 2.
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 elimination of the Power Range, Neutron Flux, Negative Rate
trip does not increase the probability or consequences of reactor
core damage accidents resulting from Rod Cluster Control Assembly
(RCCA) Misalignment events previously analyzed. The safety functions
of other safety-related systems and components have not been
altered. All other Reactor Trip System protection functions are not
impacted by the elimination of the requirement for a Power Range,
Neutron Flux, High Negative Rate trip. The Power Range, Neutron
Flux, High Negative Rate trip circuitry detects and responds to
negative reactivity insertion due to RCCA misoperation events,
should they occur. Therefore, the Power Range, Neutron Flux, High
Negative Rate trip is not assumed in the initiation of such events.
The consequences of accidents previously evaluated in the Salem
Generating Station (Salem) Updated Final Safety Analysis Report
(UFSAR) are unaffected by the proposed changes because no change to
any equipment response or accident mitigation scenario has resulted.
The proposed changes do not modify the RCCAs or change the
acceptance criteria for departure from nucleate boiling (DNB). The
TS change reflects analysis described in the UFSAR and cycle-
specific analysis performed each fuel cycle.
The proposed revisions to Salem Unit 1 Index page XII, Salem
Unit 1 TS 4.2.2.2, Salem Unit 2 TS 4.2.2.2, Salem Unit 1 TS Table
3.3-2, Salem Unit 2 SR number for boron concentration on page 3/4 9-
1, Salem Unit 1 TS 6.9.1.5.a, and Salem Unit 1 TS 6.9.1.5.b contain
changes administrative in nature that correct errors and do not
affect the intent of any TS requirements.
Therefore, the proposed changes do not involve a significant
increase in the probability or radiological 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 elimination of the Power Range, Neutron Flux, High Negative
Rate trip does not create the possibility of a new or different kind
of accident from any accident previously evaluated in the UFSAR. No
new accident scenarios, failure mechanisms, or limiting single
failures are introduced as a result of the proposed changes. The
proposed changes do not challenge the performance or integrity of
the RCCAs or any other safety-related system. The proposed changes
will have no adverse effect on the availability, operability, or
performance of the safety-related systems and components assumed to
actuate in the event of a design basis accident (DBA) or transient.
It has been demonstrated that the Power Range, Neutron Flux, High
Negative Rate trip can be eliminated by the NRC approved methodology
described in WCAP-11394-P. The Salem fuel cycle specific analyses
have confirmed that for a dropped RCCA event, no direct reactor trip
or automatic power reduction is required to meet the DNB limits for
this Condition II, ``Fault of Moderate Frequency,'' event. The Power
Range, Neutron Flux, High Negative Rate trip is not credited either
as a primary or backup mitigation feature for any other UFSAR event.
The proposed revisions to Salem Unit 1 Index page XII, Salem
Unit 1 TS 4.2.2.2, Salem Unit 2 TS 4.2.2.2, Salem Unit 1 TS Table
3.3-2, Salem Unit 2 SR number for boron concentration on page 3/4 9-
1, Salem Unit 1 TS 6.9.1.5.a, and Salem Unit 1 TS 6.9.1.5.b contain
changes administrative in nature that correct errors and do not
affect the intent of any TS requirements.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is the difference between the DNB
acceptance limit and the failure of the fuel rod cladding. The Salem
fuel cycle specific analyses have confirmed that for a dropped RCCA
event, DNB limits are not exceeded with the proposed changes.
Conformance to the licensing basis acceptance criteria for DBAs and
transients with the elimination of the Power Range, Neutron Flux,
High Negative Rate trip is demonstrated and the DNB limits are not
exceeded when the NRC approved methodology of WCAP-11394-P is
applied. The margin of safety associated with the licensing basis
acceptance criteria for any postulated accident is unchanged.
The proposed revisions to Salem Unit 1 Index page XII, Salem
Unit 1 TS 4.2.2.2, Salem Unit 2 TS 4.2.2.2, Salem Unit 1 TS Table
3.3-2, Salem Unit 2 SR number for boron concentration on page 3/4 9-
1, Salem Unit 1 TS 6.9.1.5.a, and Salem Unit 1 TS 6.9.1.5.b contain
changes administrative in nature that correct errors and do not
affect the intent of any TS requirements.
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
[[Page 40753]]
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Darrell J. Roberts.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: May 1, 2006.
Description of amendment request: The amendment would move the main
steamline discharge (safety valves and atmospheric dumps) radiation
monitors (R46) from the radiation monitoring instrumentation Technical
Specification (TS) 3.3.3.1, to the accident monitoring TS 3.3.3.7. The
purpose of the R46 monitors is to provide continuous monitoring of
high-level, post-accident releases of radioactive noble gases;
therefore, relocation to TS 3.3.3.7 is appropriate. In addition, TS
definition 1.31, ``Source Checks,'' would be modified to allow
different methods to comply with the source check requirement. This
change would affect the remaining instruments in TS 3.3.3.1, and would
allow for appropriate testing consistent with the technology of the
existing detectors, and replacement detectors in the future.
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 the R46 monitors presents no change in
the probability or the consequence of an accident, since the
monitors are used post-accident for the monitoring of high-level
releases of radioactive noble gases.
Relocation of the R46 monitors to the accident monitoring TS
3.3.3.7 is appropriate for the function of the monitors. The R46
monitors are designed to meet the requirements of NUREG-0737 Il.F.1
and the intent of RG [Regulatory Guide] 1.97. The monitor's alarm
function is used in the EOPs [Emergency Operating Procedures] to
identify a Steam Generator Tube Rupture (SGTR) event EOP entry point
and to identify which SG [steam generator] has ruptured. The
relocation of the monitor to TS 3.3.3.7 has no affect on the
function of the monitor.
The proposed change to the definition of TS 1.31 also does not
impact the accident analyses in any manner. The qualitative
assessment of monitor response will continue to be performed
verifying monitor operability.
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 relocation of the R46 monitors is primarily
administrative in nature; there will be no change in the function of
the monitors. No new accident scenarios, failure mechanisms, or
limiting single failures are introduced as a result of the proposed
changes. Post accident monitoring instrumentation is not associated
with the initiation of an accident.
The proposed change to the definition of TS 1.31 also does not
create a new or different kind of accident. The qualitative
assessment of monitor response will continue to be performed
verifying monitor operability.
3. Does the proposed change involve a significant reduction in
the margin of safety?
Response: No.
The proposed change to relocate the R46 monitors does not alter
the manner in which safety limits, limiting safety systems settings
or limiting conditions for operation are determined. The proposed
change will not alter any assumptions, initial conditions or results
specified in any accident analysis. There is no change in the R46
monitor alarm setpoint.
The proposed change to the TS definition of SOURCE CHECK does
not alter the basic requirement that a qualitative assessment of the
monitor response be performed; therefore the operability of the
monitor will continue to be verified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Darrell J. Roberts.
PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station,
Unit No. 2, Salem County, New Jersey
Date of amendment request: April 6, 2006.
Description of the amendment request: The proposed amendment
changes the existing steam generator (SG) tube surveillance program to
one that is consistent with the program proposed by the Technical
Specification Task Force (TSTF) in TSTF-449. These changes revise
Technical Specification (TS) 1.15, ``Identified Leakage,'' TS 1.21,
``Pressure Boundary Leakage,'' TS 3/4.4.6, ``Steam Generator (SG) Tube
Integrity,'' and TS 3/4.4.7.2, ``Operational Leakage,'' and add new
administrative TS 6.8.4.i, ``Steam Generator (SG) Program,'' and TS
6.9.1.10, ``Steam Generator Tube Inspection Report.'' Other editorial
changes were also made.
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 requires a Steam Generator Program that
includes performance criteria that will provide reasonable assurance
that the steam generator (SG) tubing will retain integrity over the
full range of operating conditions (including startup, operation in
the power range, hot standby, cool down 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.
The structural integrity performance criterion is:
All in-service steam generator tubes shall retain structural
integrity over the full range of normal operating conditions
(including startup, operation in the power range, hot standby, and
cool down and all anticipated transients included in the design
specification) and design basis accidents. This includes retaining a
safety factor of 3.0 against burst under normal steady state full
power operation primary-to-secondary pressure differential and a
safety factor of 1.4 against burst applied to the design basis
accident primary-to-secondary pressure differentials. Apart from the
above requirements, additional loading conditions associated with
the design basis accidents, or combination of accidents in
accordance with the design and licensing basis, shall also be
evaluated to determine if the associated loads contribute
significantly to burst or collapse. In the assessment of tube
integrity, those loads that do significantly affect burst or
collapse shall be determined and assessed in combination with the
loads due to pressure with a safety factor of 1.2 on the combined
primary loads and 1.0 on axial secondary loads.
The accident induced leakage performance criterion is:
The primary-to-secondary accident induced leakage rate for any
design basis accidents, other than a SG tube rupture, shall not
exceed the leakage rate assumed in the accident analysis in terms of
total leakage rate for all SGs and leakage rate for an individual
SG. Leakage is not to exceed 1 gpm [gallon per minute] per SG.
The operational leakage performance criterion is:
The reactor coolant system operational primary-to-secondary
leakage through any
[[Page 40754]]
one SG shall be limited to 150 gallons per day.
A steam generator tube rupture (SGTR) event is one of the design
basis accidents that are analyzed as part of a plant's licensing
basis. In the analysis of an 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 main steam line break
(MSLB), rod ejection, and reactor coolant pump locked rotor, the
tubes are assumed to retain their structural integrity (i.e., they
are assumed not to rupture). These analyses 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 retained 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 as part of these TS changes
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 Steam Generator Program required by the
proposed addition of TS 6.8.4.i. The program defined by NEI [Nuclear
Energy Institute] 97-06 includes a framework that incorporates a
balance of prevention, inspection, evaluation, repair, and leakage
monitoring.
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 Salem TS 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 through 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 that allows SR [Surveillance Requirement]
4.4.7.2.1.d to not be performed until 12 hours after establishment
of steady state operation is consistent with NUREG 1431, ``Standard
Technical Specifications, Westinghouse Plants'', and ensures the
surveillance requirement is appropriate for the LCO [Limiting
Condition for Operation].
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 TS 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 TS.
Therefore, the proposed changes do not affect the consequences
of an SGTR accident and the probability of such an accident is
reduced.
In addition, the proposed changes do not affect the
probabilities or consequences of an MSLB, rod ejection, or a reactor
coolant pump locked rotor event.
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 performance based requirements are an improvement
over the requirements imposed by the current TS.
Implementation of the proposed Steam Generator 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 Steam Generator 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 changes do 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.
The proposed change that allows SR 4.4.7.2.1.d to not be
performed until 12 hours after establishment of steady state
operation is consistent with NUREG 1431, ``Standard Technical
Specifications, Westinghouse Plants'', and ensures the surveillance
requirement is appropriate for the LCO.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The 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
also isolate the radioactive fission products in the primary coolant
from the secondary system. In summary, the safety function of a SG
is maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the Steam Generator Program to manage SG
tube inspection, assessment, repair and plugging. The requirements
established by the Steam Generator Program are consistent with those
in the applicable design codes and standards and are an improvement
over the requirements in the current TS.
The proposed change that allows SR 4.4.7.2.1.d to not be
performed until 12 hours after establishment of steady state
operation is consistent with NUREG 1431, ``Standard Technical
Specifications, Westinghouse Plants'', and ensures the surveillance
requirement is appropriate for the LCO.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed changes to the
TS.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Darrell J. Roberts.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: June 2, 2006.
Description of amendment requests: The amendment proposes to revise
Technical Specification (TS) 3.8.1, ``AC [alternating current]
Sources--Operating,'' and TS 3.8.3, ``Diesel Fuel Oil, Lube Oil, and
Starting Air,'' to increase the required amount of stored diesel fuel
oil to support a change to Ultra Low Sulfur Diesel fuel from California
diesel fuel presently in use. This change in the type of fuel oil is
mandated by California air pollution control regulations.
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.
This proposed change increases the minimum amount of stored
diesel fuel. The change supports the use of Ultra Low Sulfur Diesel
(ULSD) fuel rather than the existing California Air Resources Board
diesel fuel as mandated by California air pollution control
regulations (Title 13 California Code of
[[Page 40755]]
Regulations Division 3, Chapter 5, Article 2, Sections 2280-2285).
Technical Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube Oil,
and Starting Air,'' requires that each diesel generator have
sufficient fuel to operate for a period of 7 days, while the diesel
generator (DG) is supplying maximum post Loss of Coolant Accident
(LOCA) load demand.
Because the Lower Heating Value (LHV) per gallon of ULSD fuel is
less than that of existing diesel fuel, it was necessary to re-
calculate the amount of fuel required to supply necessary loads for
the required time periods. For Modes 1 through 4, the resulting
minimum volumes of ULSD fuel are 48,400 gallons and 41,800 gallons
for the 7-day and 6-day fuel supply, respectively. For Modes 5 and
6, the required volumes of ULSD fuel are 43,600 gallons and 37,400
gallons for a 7-day supply and a 6-day supply, respectively.
The DGs and the associated support systems such as the fuel oil
storage and transfer systems are designed to mitigate accidents and
are not accident initiators. Increasing the minimum volumes of
stored fuel in the storage and day tanks will not result in a
significant increase in the probability of any accident previously
evaluated.
Following implementation of this proposed change, there will be
no change in the ability of the diesel generators to supply maximum
post-LOCA load demand for 7 days. The proposed minimum volumes of
fuel, 48,400 gallons and 41,800 gallons, ensure that a 7-day and [a]
6-day supply of fuel, respectively, are available in Modes 1 through
4. The proposed minimum volumes of fuel, 43,600 gallons and 37,400
gallons, ensure that a 7-day and a 6-day supply, respectively, of
fuel is available in Modes 5 and 6. This is identical to the current
requirements, except for the increased volume of fuel required due
to the decreased heat content of the ULSD fuel. Therefore, this
change will not result in a significant increase in the consequences
of any accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Following this change, the diesel generators will still be able
to supply maximum post-LOCA load demand. The current 7-day and 6-day
fuel supply requirements will be maintained following this change.
The new required fuel oil volumes are within the capacities of the
fuel oil storage tanks.
Therefore, this proposed change will not create the possibility
of a new or different kind of accident from any accident that has
been previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The Bases to TS 3.8.3 state that ``[e]ach diesel generator (DG)
is provided with a storage tank having a fuel oil capacity
sufficient to operate that diesel for a period of 7 days, while the
DG is supplying maximum post loss of coolant accident load demand.''
When the fuel oil tank level is less than required to support the 7-
day of operation, the required action depends on whether or not a 6-
day supply of fuel is available.
The proposed tank level limits will maintain these 7-day and 6-
day fuel supply requirements in all operating Modes following
changeout to ULSD fuel.
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 requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Branch Chief: David Terao.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: May 25, 2006.
Description of amendment request: The amendment would revise the
Technical Specifications (TSs) to adopt NRC-approved Revision 4 to
Technical Specification Task Force (TSTF) Standard Technical
Specification Change Traveler TSTF-372, ``Addition of LCO [Limiting
Condition for Operation] 3.0.8, Inoperability of Snubbers.'' The
amendment would add (1) a new LCO 3.0.8 addressing when one or more
required snubbers are unable to perform their associated support
function(s) (i.e., the snubber is inoperable) and (2) a reference to
LCO 3.0.8 in LCO 3.0.1 on when LCOs shall be met.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on November 24, 2004 (69 FR 68412), on possible
license amendments adopting TSTF-372 using the NRC's consolidated line
item improvement process (CLIIP) for amending licensee's TSs, which
included a model safety evaluation (SE) and model no significant
hazards consideration (NSHC) determination. 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), which included the resolution of public comments on the
model SE. The May 4, 2005, notice of availability referenced the
November 24, 2004, notice. The licensee has affirmed the applicability
of the following NSHC determination in its application.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--Does the proposed change involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The proposed 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. The
postulated seismic event requiring snubbers is a low-probability
occurrence and the overall TS system safety function would still be
available for the vast majority of anticipated challenges.
Therefore, the probability of an accident previously evaluated is
not significantly increased, if at all. The consequences of an
accident while relying on allowance provided by proposed LCO 3.0.8
are no different than the consequences of an accident while relying
on the TS required actions in effect without the allowance provided
by proposed LCO 3.0.8. Therefore, the consequences of an accident
previously evaluated are not significantly affected by this change.
The addition of a requirement to assess and manage the risk
introduced by this change will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or 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?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Allowing delay times for entering [a] supported system TS when
inoperability is due solely to inoperable snubbers, if risk is
assessed and managed, will not introduce new failure modes or
effects and will not, in the absence of other unrelated failures,
lead to an accident whose consequences exceed the consequences of
accidents previously evaluated. The addition of a requirement to
assess and manage the risk introduced by this change will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
Criterion 3--Does the proposed change involve a significant
reduction in the margin of safety?
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. The risk impact of the
proposed TS changes was assessed following
[[Page 40756]]
the three-tiered approach recommended in [NRC] RG [Regulatory Guide]
1.177. A bounding risk assessment was performed to justify the
proposed TS changes. This application of LCO 3.0.8 is predicated
upon the licensee's performance of a risk assessment and the
management of plant risk [, which is required by the proposed TS
3.0.8]. The net change to the margin of safety is insignificant.
Therefore, this change does not involve a significant reduction in a
margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: David Terao.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: May 25, 2006.
Description of amendment request: The amendment would revise
Technical Specifications 3.1.7, ``Rod Position Indication,'' 3.2.1,
``Heat Flux Hot Channel Factor (FCQ(Z)) (FQ
Methodology),'' 3.2.4, ``Quadrant Power Tilt Ratio (QPTR),'' and 3.3.1,
``Reactor Trip System (RTS) Instrumentation.'' The proposed changes are
to allow use of the Westinghouse proprietary computer code, the Best
Estimate Analyzer for Core Operations--Nuclear (BEACON). The new BEACON
power distribution monitoring system (PDMS) would augment the
functional capability of the neutron flux mapping system for the
purposes of power distribution surveillances at the Callaway Plant.
Certain required actions, for when a limiting condition for operation
is not met, and certain surveillance requirements are being changed to
refer to power distribution measurements or measurement information of
the core.
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 PDMS performs continuous core power distribution monitoring
with data input from existing plant instrumentation. This system
utilizes an NRC-approved Westinghouse proprietary computer code,
i.e., Best Estimate Analyzer for Core Operations [mu] Nuclear
(BEACON), to provide data reduction for incore flux maps, core
parameter analysis, load follow operation simulation, and core
predication. The PDMS does not provide any protection or control
system function. Fission product barriers are not impacted by these
proposed changes. The proposed changes occurring with PDMS will not
result in any additional challenges to plant equipment that could
increase the probability of any previously evaluated accident. The
changes associated with the PDMS do not affect plant systems such
that their function in the control of radiological consequences is
adversely affected. These proposed changes will therefore not affect
the mitigation of the radiological consequences of any accident
described in the Final Safety Analysis Report (FSAR) [for the
Callaway Plant].
Use of the PDMS supports maintaining the core power distribution
within required limits. Further continuous on-line monitoring
through the use of PDMS provides significantly more information
about the power distributions present in the core than is currently
available. This results in more time (i.e., earlier determination of
an adverse condition developing) for operation action prior to
having an adverse condition develop that could lead to an accident
condition or to unfavorable initial conditions for an accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do[es] the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Other than use of the PDMS to monitor core power distribution,
implementation of the PDMS and associated Technical Specification
changes has no impact on plant operations or safety, nor does it
contribute in any way to the probability or consequences of an
accident. No safety-related equipment, safety function, or plant
operation [other than core power distribution monitoring] will be
altered as a result of this proposed change. The possibility for a
new or different type of accident from any accident previously
evaluated is not created since the changes associated with [the]
implementation of the PDMS do not result in a change to the design
basis of any plant component or system [other than to the PDMS]. The
evaluation of the effects of using the PDMS to monitor core power
distribution parameters shows that all design standards and
applicable safety criteria limits are met. [The PDMS is to monitor
the core power distribution and is, therefore, not an accident
initiator.]
The proposed changes do not result in any event previously
deemed incredible being made credible [by the implementation of the
PDMS]. Implementation of the PDMS will not result in any additional
adverse condition and will not result in any increase in the
challenges to safety systems. The cycle-specific variables required
by the PDMS are calculated using NRC-approved methods. The Technical
Specifications will continue to require operation within the
required core operating limits, and appropriate actions will
continue to be [required to be] taken when or if limits are
exceeded.
The proposed change, therefore, does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do[es] the proposed change involve a significant reduction in
a margin of safety?
Response: No.
No margin of safety is adversely affected by the implementation
of the PDMS. The margins of safety provided by [the] current
Technical Specification requirements and limits remain unchanged, as
the Technical Specifications will continue to require operation
within the core limits that are based on NRC-approved reload design
methodologies. [These NRC-approved reload design methodologies are
not being changed.] Appropriate measures exist to control the values
of these cycle-specific limits, and appropriate actions will
continue to be specified and [required to be] taken for when limits
are violated. Such actions remain unchanged.
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: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: David Terao.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: June 2, 2006.
Description of amendment request: The amendment would revise
Surveillance Requirement 3.5.2.8 in the Technical Specifications by
replacing the phrase ``trash racks and screens'' with the word
``strainers.'' The amendment reflects the replacement of the
containment sump suction inlet trash racks and screens with a complex
strainer design with significantly larger effective area in the
upcoming Refueling Outage 15. This is in response to Generic Letter
2004-02, ``Potential Impact of Debris Blockage on Emergency
Recirculation during Design Basis Accidents at Pressurized-Water
Reactors,'' dated September 13, 2004.
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
[[Page 40757]]
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 consequences of accidents evaluated in the Updated Safety
Analysis Report (USAR) [for the Wolf Creek Generating Station] that
could be affected by the proposed change are those involving the
pressurization of containment and associated flooding of the
containment and recirculation of this fluid within the Emergency
Core Cooling System (ECCS) or the Containment Spray System (CSS)
(e.g., Loss of Coolant Accidents). The proposed change does not
impact the initiation or probability of occurrence of any accident.
[The containment sump trash racks and screens, and the sump
strainers that are replacing the trash racks and screens are not
initiators of accidents.]
Although the configurations of the existing containment
recirculation sump trash racks and screen[s,] and the replacement
sump strainer assemblies are different, they serve the same
fundamental purpose of passively removing debris from the sump's
suction supply of the supported system pumps. Removal of trash racks
does not impact the adequacy of the pump NPSH [net positive suction
head] assumed in the safety analysis. Likewise, the change does not
reduce the reliability of any supported systems or introduce any new
system interactions. The greatly increased surface area of the new
strainer is designed to reduce head loss [at the containment sump]
and reduce the approach velocity at the strainer face significantly,
decreasing the risk of impact from large debris entrained in the
sump flow stream.
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 containment recirculation sump strainers are a passive
system used for accident mitigation. As such, they cannot be
accident initiators. Therefore, there is no possibility that this
change could create any new or different kind of accident.
No new accident scenarios, transient precursors, or limiting
single failures are introduced as a result of the proposed change.
There will be no adverse effect[s] or challenges imposed on any
safety related system as a result of the change. Therefore, the
possibility of a new or different type of accident is not created.
[The containment recirculation sump suction inlet trash racks and
screens are being replaced with a complex strainer design with
significantly larger effective surface area to reduce head loss and
reduce the approach velocity at the strainer face significantly,
decreasing the risk of impact from large debris entrained in the
sump flow stream.]
There are no changes which would cause the malfunction of safety
related equipment, assumed to be OPERABLE in the accident analyses,
as a result of the proposed Technical Specification change. No new
equipment performance burdens are imposed. The possibility of a
malfunction of safety related equipment with a different result [or
consequences] is not created.
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 does not affect the acceptance criteria for
any analyzed event nor is there a change to any Safety Analysis
Limit (SAL). There will be no effect on the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined nor will there be any effect on those plant
systems necessary to assure the accomplishment of protection
functions. The proposed change does not adversely affect the fuel,
fuel cladding, Reactor Coolant System, or containment integrity. The
radiological dose consequence acceptance criteria listed in the
Standard Review Plan [for accidents] will continue to be met.
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: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: David Terao.
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].
Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423,
Millstone Power Station, Unit Nos. 2 and 3, New London County,
Connecticut
Date of application for amendments: March 9, 2005, as supplemented
by letter dated July 7, 2005.
Brief description of amendments: The amendments revised the
Millstone Power Station, Unit Nos. 2 and 3 Technical Specifications to
incorporate wording related to the reactor coolant system, electrical
power system and refueling operations to provide operational
flexibility during mode changes or addition of coolant during shutdown
operations.
Date of issuance: June 28, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 293 and 230.
Facility Operating License Nos. DPR-65 and NPF-49: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 24, 2005 (70 FR
29788). The additional information provided in
[[Page 40758]]
the supplemental letter dated July 7, 2005, did not expand the scope of
the application as noticed and did not change the NRC staff's original
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 28, 2006.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: January 5, 2005, as supplemented
November 21, 2005.
Brief description of amendments: The amendments revised Technical
Specifications (TSs) 5.5.19.b, 5.1.19.c, and TS Surveillance
Requirement (SR) 3.8.1.9 associated with the Lee Combustion Turbine
(LCT) testing program. TS 5.5.19 required verification that an LCT can
supply the equivalent of one unit's maximum safeguards loads, plus two
units' Mode 3 loads when connected to the system grid every 12 months.
The amendments clarified this requirement as ``Verify an LCT can supply
equivalent of one unit's Loss of Coolant Accident (LOCA) loads plus two
units' Loss of Offsite Power (LOOP) loads when connected to system grid
every 12 months.'' TS 5.5.19.c and SR 3.8.1.9 were revised for
consistency.
Date of Issuance: July 5, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 352/354/353.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: February 15, 2005 (70
FR 7764). The additional information provided in the supplemental
letter dated November 21, 2005, did not expand the scope of the
application as noticed and did not change the NRC staff's original
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 5, 2006.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: December 29, 2005.
Brief description of amendment: The amendment deleted License
Condition, Section 2.F, that requires the reporting of violations in
Section 2.C of the Facility Operating License.
Date of issuance: June 28, 2006.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 116.
Facility Operating License No. NPF-69: Amendment revised the
Facility Operating License.
Date of initial notice in Federal Register: April 25, 2006 (71 FR
23958).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 28, 2006.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia
Date of application for amendments: February 17, 2006.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) adding Limiting Condition for Operation
(LCO) 3.0.8 to allow a delay time for entering a supported system 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).
Date of issuance: June 29, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 250/194.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the licenses and the technical specifications.
Date of initial notice in Federal Register: April 25, 2006 (71 FR
23960).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 29, 2006.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Unit Nos. 1 and 2, Burke County,
Georgia
Date of application for amendments: February 17, 2006.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) adding Limiting Condition for Operation
(LCO) 3.0.8 and renumbering existing LCO 3.0.8 to LCO 3.0.9 to allow a
delay time for entering a supported system 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).
Date of issuance: June 29, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 141/121.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Licenses and the Technical Specifications.
Date of initial notice in Federal Register: April 25, 2006 (71 FR
23960).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 29, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: December 19, 2005, as
supplemented by letter dated March 30, 2006.
Brief description of amendments: The amendments modified several
parts of Technical Specification Surveillance Requirement (SR) 4.0.5,
both to change the surveillance intervals for which the 25 percent
extension provided in SR 3.0.2 would apply, and to replace the
references in SR 4.0.5 to the American Society of Mechanical Engineers
(ASME) Boiler and Pressure Vessel Code, Section XI, with the ASME
Operation and Maintenance Code.
Date of issuance: June 16, 2006.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 308 and 297.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the technical specifications.
Date of initial notice in Federal Register: February 14, 2006 (71
FR 7183).
The supplemental letter dated March 30, 2006, provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 16, 2006.
No significant hazards consideration comments received: No.
[[Page 40759]]
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of application for amendments: April 20, 2006, as supplemented
on May 15, 2006.
Brief description of amendments: These amendments revised the
reactor coolant pressure and temperature limits, low-temperature
overpressure protection system (LTOPS) setpoint values, and LTOPS
enable temperatures for up to 28.8 effective full-power years (EFPYs)
and 29.4 EFPYs of operation at Surry Power Station, Unit Nos. 1 and 2,
respectively.
Date of issuance: June 29, 2006.
Effective date: As of the date of issuance.
Amendment Nos.: 248/247.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments revised the License and the Technical Specifications.
Date of initial notice in Federal Register: April 28, 2006 (71 FR
25249).
The May 15, 2006, supplement contained clarifying information only
and did not change the initial proposed no significant hazards
consideration determination or expand the scope of the initial
application.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 29, 2006.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 11th day of July.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 06-6246 Filed 7-17-06; 8:45 am]
BILLING CODE 7590-01-P