[Federal Register Volume 70, Number 99 (Tuesday, May 24, 2005)]
[Notices]
[Pages 29785-29808]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-10063]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from April 29, 2005 through May 12, 2005. The
last biweekly notice was published on May 10, 2005 (70 FR 24645).
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
[[Page 29786]]
must also provide references to those specific sources and documents of
which the petitioner is aware and on which the petitioner/requestor
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or fact.
Contentions shall be limited to matters within the scope of the
amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-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].
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: January 21, 2005.
Description of amendment request: The proposed amendment would
implement the Alternative Source Term (AST) for the analysis of the
radiological consequences of a design-basis Loss-of-Coolant Accident
(LOCA). There are no changes proposed to the Operating License or
Technical Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The Proposed Change Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated
Revision of the LOCA analysis to the Alternative Source Term
methodology does not affect the design or operation of HBRSEP [H. B.
Robinson Steam Electric Plant], Unit No. 2. Rather, once the
occurrence of an accident has been postulated, the new source term
is an input to evaluate the consequences of the postulated accident.
The implementation of the Alternative Source Term has been evaluated
in revisions to the LOCA dose analysis at HBRSEP, Unit No. 2. Based
on the results of this analysis, it has been demonstrated that the
dose consequences are within the regulatory guidance provided by the
NRC. This guidance is presented in 10 CFR 50.67 and Regulatory Guide
1.183.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. The Proposed Change Does Not Create the Possibility of a New or
Different Kind of Accident From Any Previously Evaluated
The proposed change does not affect plant structures, systems,
or components. The proposed change is to an evaluation methodology
and does not initiate design basis accidents.
Thus, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The Proposed Change Does Not Involve a Significant Reduction in the
Margin of Safety
The proposed change is associated with the implementation of a
new licensing basis for HBRSEP, Unit No. 2. The new licensing basis
implements an Alternative Source Term in accordance with 10 CFR
50.67 and the associated Regulatory Guide 1.183. The results of the
revised limiting design basis analysis are subject to revised
acceptance criteria. The analysis has been performed using
conservative methodologies in accordance with regulatory guidance or
other methodologies approved by the NRC in prior plant-specific
license amendments. The dose consequences are within the acceptance
criteria found in the regulatory guidance associated with
Alternative Source Terms.
The proposed change continues to ensure that doses at the
exclusion area and low population zone boundaries, as well as the
control room, are within the corresponding regulatory limits.
Specifically, the margin of safety for the radiological consequences
of these accidents is considered to be that provided by meeting the
applicable regulatory limits.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
[[Page 29787]]
NRC Section Chief: Michael L. Marshall, Jr.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: February 14, 2005.
Description of amendment request: The proposed amendment would
revise the surveillance requirements (SRs) for the station batteries as
specified in Technical Specification (TS) SR 3.8.4.5, the battery
service test, and TS SR 3.8.4.6, the battery performance test.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the Proposed Changes Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated?
No. The proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed surveillance changes will continue to ensure
that the DC system is tested in a manner that will verify
operability. Performance of the required system surveillances, in
conjunction with the applicable operational and design requirements
for the DC system, provide assurance that the system will be capable
of performing the required design functions for accident mitigation
and also that the system will perform in accordance with the
functional requirements for the system as described in the Updated
Final Safety Analysis Report for HBRSEP [H. B. Robinson Steam
Electric Plant], Unit No. 2. This ensures that the rate of
occurrence and consequences of analyzed accidents will not change.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the Proposed Changes Create the Possibility of a New or
Different Kind of Accident From Any Previously Evaluated?
No. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated. The
proposed surveillance requirement changes will continue to ensure
that the DC system is tested in a manner that will verify
operability. No physical changes to the HBRSEP, Unit No. 2, systems,
structures, or components are being implemented. There are no new or
different accident initiators or sequences being created by the
proposed Technical Specifications changes. Therefore, these changes
do not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Do the Proposed Changes Involve a Significant Reduction in
the Margin of Safety?
No. The proposed changes do not involve a significant reduction
in the margin of safety. The proposed DC system surveillance
requirement changes provide appropriate and applicable surveillances
for the DC system. The proposed changes to surveillance requirements
for the DC system will continue to ensure system operability.
Therefore, these changes do not affect any margin of safety for
HBRSEP, Unit No. 2.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Section Chief: Michael L. Marshall, Jr.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: March 3, 2005.
Description of amendment request: The proposed amendment would
revise the requirements of Technical Specification (TS) 5.6.5, ``Core
Operating Limits Report (COLR).'' Specifically, the proposed change
would add topical report EMF-2103(P)(A), ``Realistic Large Break LOCA
[loss-of-coolant accident] Methodology for Pressurized Water
Reactors,'' to the list of documents specified in TS 5.6.5. TS 5.6.5
lists the approved methodologies that can be used to determine the core
operating limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The Proposed Change Does Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously
Evaluated?
The proposed methodology will be reviewed and approved by the
NRC prior to its use for HBRSEP [H. B. Robinson Steam Electric
Plant], Unit No. 2. Analyzed events are assumed to be initiated by
the failure of plant structures, systems, or components. The
determination of core operating limits in accordance with this new
methodology will meet the limitations specified in the NRC safety
evaluation of the new methodology. The topical report associated
with the new methodology demonstrates that the integrity of the fuel
will be maintained and that design requirements will continue to be
met. The proposed change does not involve physical changes to any
plant structure, system, or component. Therefore, the probability of
occurrence for a previously analyzed accident is not significantly
increased.
The consequences of a previously analyzed accident are dependent
on the initial conditions assumed for the analysis, the behavior of
the fuel during the analyzed accident, the availability and
successful functioning of the equipment assumed to operate in
response to the analyzed event, and the setpoints at which these
actions are initiated. The proposed methodology continues to meet
applicable design and safety analyses acceptance criteria. The
proposed change does not affect the performance of any equipment
used to mitigate the consequences of an analyzed accident. As a
result, no analysis assumptions are violated and there are no
adverse effects on the factors that contribute to offsite or onsite
dose as the result of an accident. The proposed change does not
affect setpoints that initiate protective or mitigative actions. The
proposed change ensures that plant structures, systems, or
components are maintained consistent with the safety analysis and
licensing bases. Based on this evaluation, there is no significant
increase in the consequences of a previously analyzed event.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. The Proposed Change Does Not Create the Possibility of a New
or Different Kind of Accident From Any Previously Evaluated?
The proposed change does not involve any physical alteration of
plant systems, structures, or components, other than allowing for
fuel design in accordance with NRC approved methodologies. The
proposed methodology continues to meet applicable criteria for Large
Break Loss of Coolant Accident (LBLOCA) analysis. No new or
different equipment is being installed. No installed equipment is
being operated in a different manner. There is no alteration to the
parameters within which the plant is normally operated or in the
setpoints that initiate protective or mitigative actions. As a
result, no new failure modes are being introduced. There are no
changes in the methods governing normal plant operation, nor are the
methods utilized to respond to plant transients altered. Therefore,
the proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The Proposed Change Does Not Involve a Significant Reduction
in the Margin of Safety?
The margin of safety is established through the design of the
plant structures, systems, and components, through the parameters
within which the plant is operated, through the establishment of the
setpoints for the actuation of equipment relied upon to respond to
an event, and through margins contained within the safety analyses.
The proposed change in the methodology used for LBLOCA analyses does
not impact the condition or performance of structures, systems,
setpoints, and components relied upon for accident mitigation. The
proposed
[[Page 29788]]
change does not significantly impact any safety analysis assumptions
or results. Therefore, the proposed change does not result in 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: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Section Chief: Michael L. Marshall, Jr.
Dominion Nuclear Connecticut, Inc., Docket Nos. 50-245, 50-336, and 50-
423, Millstone Power Station, Unit Nos. 1, 2, and 3, New London County,
Connecticut
Date of amendment request: December 21, 2004.
Description of amendment request: The requested change will delete
Technical Specification (TS) requirements for annual Occupational
Radiation Exposure Reports (all units), annual report regarding
challenges to pressurizer relief and safety valves (Units 2 and 3), and
Monthly Operating Reports (Units 2 and 3).
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
license amendment applications in the Federal Register on June 23, 2004
(69 FR 35067). The licensee affirmed the applicability of the model
NSHC determination in its application dated December 21, 2004.
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:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the TSs reporting requirements to
provide a monthly operating letter report of shutdown experience and
operating statistics if the equivalent data is submitted using an
industry electronic database. It also eliminates the TS reporting
requirement for an annual occupational radiation exposure report,
which provides information beyond that specified in NRC regulations.
The proposed change involves no changes to plant systems or accident
analyses. As such, the change is administrative in nature and does
not affect initiators of analyzed events or assumed mitigation of
accidents or transients. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
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.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve significance hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Section Chief: Darrell J. Roberts.
Dominion Nuclear Connecticut Inc., et al., Docket No. 50-336, Millstone
Power Station, Unit No. 2, New London County, Connecticut
Date of amendment request: March 9, 2005.
Description of amendment request: Current Technical Specifications
(TSs) require that all operations involving a reduction in reactor
coolant boron concentration or that involve positive reactivity changes
be suspended under certain conditions. The requested changes modify the
TSs to incorporate wording related to the reactor coolant system (RCS),
electrical power systems, and refueling operations to provide
operational flexibility during mode changes or addition of coolant
during shutdown operations. Additionally, changes are to be made to the
TS bases, as appropriate.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1: Does the proposed amendment involve a significant increase
in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not in any way alter the SDM [shutdown
margin] or refueling boron concentration. It limits introduction of
coolant into the RCS of reactivity more positive than that necessary
to meet the required SDM or refueling boron concentration. This
proposed change does not affect the input or assumptions for any
accidents previously evaluated nor does it affect initiation of an
accident. Based on this discussion, the proposed amendment does not
increase the probability or consequence of an accident previously
evaluated.
Criterion 2: Does the proposed amendment create the possibility of a
new or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change allows introduction of coolant into the RCS
with different temperature or lower boron concentration, however,
the required boron concentration or SDM is maintained. The proposed
amendment does not introduce failure modes, accident initiators, or
malfunctions that would cause a new or different kind of accident.
No plant modifications are associated with the change. Therefore,
the proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Criterion 3: Does the proposed amendment involve a significant
reduction in a margin of safety?
Response: No.
The proposed change provides the flexibility necessary for
continued safe reactor operations while limiting any potential for
excess positive reactivity additions. [The] SDM and required boron
concentration are not affected. Therefore, based on the above, the
proposed amendment 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: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
NRC Section Chief: Darrell J. Roberts.
Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone
Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: December 23, 2004.
Description of amendment request: The requested amendment would
[[Page 29789]]
relocate certain Technical Specifications regarding refueling
operations to the Technical Requirements Manual (TRM).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1: Does the proposed amendment involve a significant increase
in the probability or consequences of an accident previously evaluated?
Response: No.
The communications equipment, refueling machine, and spent fuel
pool crane are not designed to perform accident mitigation
functions. The proposed change to relocate selected refueling
specifications does not modify any plant equipment and does not
impact any failure modes that could lead to an accident. Relocating
the specifications to the TRM where changes would be controlled
under the 10 CFR 50.59 process does not change the ability of the
communications or refueling equipment to function as expected.
Additionally, these specifications have no affect on the consequence
of any analyzed accident since the equipment is not related to
accident mitigation. Based on this discussion, the proposed
amendment does not increase the probability or consequences of an
accident previously evaluated.
Criterion 2: Does the proposed amendment create the possibility of a
new or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change[s] do[es] not modify any plant equipment and
there is no impact on the capability of the existing equipment to
perform their intended functions to move fuel safely or conduct
refueling operations while in contact with the control room. No
system setpoints are being modified and no changes are being made to
the method in which refueling operations are conducted. No changes
to the heavy loads program are being proposed by this change. No new
failure modes are introduced by the proposed changes. The proposed
amendment does not introduce accident initiators or malfunctions
that would cause a new or different kind of accident. Therefore, the
proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Criterion 3: Does the proposed amendment involve a significant
reduction in a margin of safety?
Response: No.
The relocation of Technical Specification 3/4.9.5, ``Refueling
Operations, Communications,'' to the TRM does not imply any
reduction in its importance in [e]nsuring communication between the
control room and the refueling station. The proposed change will not
alter the requirement on communication between the control room and
the refueling station, it will not alter any of the assumptions used
in the fuel handling accident analysis, nor will it cause any safety
system parameters to exceed their acceptance limit. The relocation
of Technical Specification 3/4.9.6, ``Refueling Machine'' to the TRM
does not alter the requirement for the lifting device on the
refueling machine to have adequate capacity or for the interlocks to
be demonstrated operable prior to fuel movement. The assumptions
used in the accident analysis are not impacted by this change and no
impact to any safety system parameters will result. The relocation
of Technical Specification 3/4.9.7, ``Crane Travel--Spent Fuel
Storage Areas,'' to the TRM will not alter the requirement that the
crane interlocks and/or physical stops are operable, nor will it
alter any of the assumptions used in the fuel handling accident
analysis. Heavy load lifts are administratively controlled by a safe
load path and crane interlocks. The proposed change[s] do[es] not
modify any heavy load path criteria. Administrative changes
associated with the proposed revision such as relocation of
associated Technical Specification Bases to the TRM will not have an
impact on any established safety margins.
The proposed change[s] do[es] not affect any of the assumptions
used in the accident analysis, nor do they affect any operability
requirements for equipment important to plant safety. Therefore, the
proposed change[s] will not result in a significant reduction in the
margin of safety as defined in the Bases for Technical
Specifications covered in this License Amendment Request.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
NRC Section Chief: Darrell J. Roberts.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: November 25, 2002, as supplemented by
letters dated November 13, and December 16, 2003, September 22, 2004,
and April 6, 2005.
Description of amendment request: The amendments would revise the
Technical Specifications (TS) for the Ventilation Filter Testing
Program (VFTP), Annulus Ventilation System (AVS), Auxiliary Building
Filtered Ventilation Exhaust System (ABFVES), Fuel Handling Ventilation
Exhaust System (FHVES), and Control Room Area Ventilation System
(CRAVS), and containment penetrations.
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:
First Standard
Does operation of the facility in accordance with the proposed
amendment involve a significant increase in the probability or
consequences of an accident previously evaluated? No.
This license amendment request proposes amendments to the system
TS and/or Bases and/or VFTP TS requirements for the AVS, ABFVES,
FHVES, and CRAVS. It also proposes amendments to the TS and Bases
for Containment Penetrations. The AVS is in standby during normal
plant operations and operates only following a Safety Injection
signal or during a test. It is not an accident initiator. The ABFVES
is in operation during normal plant operations. However, the ABFVES
is not used in direct support of any phase of power generation or
conversion or transmission, shutdown cooling, fuel handling
operations, or processing of radioactive fluids. Therefore, it is
not an accident initiator. The FHVES is utilized to support fuel
handling operations when moving recently irradiated fuel. It is not
an accident initiator. The CRAVS operates during normal plant
operations. However, it is not an accident initiator (the CRAVS
being defined so as to exclude equipment that maintains an
appropriately low temperature in the control room). The status of
containment penetrations is required to be controlled so as to
minimize the consequences of a fuel handling accident or a weir gate
drop accident. The containment penetrations by themselves are not
accident initiators. No accident initiators are associated with the
changes proposed in this license amendment request. For these
reasons, operation of the facility in accordance with this proposed
amendment does not involve a significant increase in the probability
of any accident previously evaluated.
In support of the proposed amendment, an analysis has been
performed to determine the radiological consequences of the design
basis [Loss of Coolant Accident] LOCA at Catawba Nuclear Station.
The analysis made use of the Alternative Source Term (AST)
methodology and in general conformed to the regulatory positions of
Regulatory Guide 1.183 and the draft regulatory positions of DG-
1111. Total Effective Dose Equivalent (TEDE) radiation doses at the
Exclusion Area Boundary (EAB), boundary of the Low Population Zone
(LPZ), and to the control room operators were calculated and found
to be acceptable. TEDEs were calculated for a design basis LOCA
postulated for a Catawba nuclear unit operating with all low
enriched uranium (LEU) fuel and with 4 mixed oxide (MOX) lead fuel
assemblies (LFAs). It was found that insertion of 4 MOX LFAs did not
produce a significant increase in the TEDEs for a design basis LOCA.
* * * * *
The new value for the control room TEDE radiation dose is higher
than the TEDE radiation dose equivalent to the radiation
[[Page 29790]]
doses currently reported in the UFSAR. However, the limiting control
room TEDE radiation dose reported in this submittal is lower than
the acceptance criterion * * * The new LPZ TEDE radiation dose is
higher than the equivalent TEDE radiation dose currently
represented. On the other hand, the margin to the acceptance
criterion is [large] * * *. The TEDE radiation doses newly computed
at the EAB for the design basis LOCA are lower than the
corresponding equivalent EAB TEDE radiation dose currently
represented in the UFSAR. The margin in the EAB TEDE radiation dose
to the guideline value is [also large]. * * * In all cases, there is
significant margin between the newly calculated post-LOCA TEDE
radiation doses and the corresponding regulatory guideline values.
In the sense that the margins to the germane regulatory guideline
values are still large, the new values of TEDE radiation doses are
comparable to the equivalent TEDE associated with the post-LOCA
radiation doses currently listed in the UFSAR. Furthermore, these
margins for the design basis LOCA do not significantly decrease with
insertion of the 4 MOX LFAs. Therefore, the proposed amendment is
determined to not result in a significant increase in accident
consequences.
AST analyses also were completed for the design basis locked
rotor accident (LRA) and rod ejection accident (REA). Again, these
design basis accidents were postulated to occur at a Catawba nuclear
unit operating with either an all LEU core or with 4 MOX LFAs. The
TEDEs following these design basis accidents were compared to the
equivalent TEDEs associated with the current license basis analyses.
The equivalent TEDEs were computed from the post-accident whole body
and thyroid radiation doses using the method prescribed in
Regulatory Guide 1.183, as noted above. TEDEs only at offsite
locations were compared as post-accident control room radiation
doses are not reported for these design basis accidents in the
Catawba UFSAR.
* * * * * * *
For the EAB, LPZ, and control room, the post-LRA TEDEs are seen
to increase from the values equivalent to the radiation doses from
the current license basis analyses. (This is attributed primarily to
the increase in assumed fraction of the fuel pins with clad failure
following a design basis LRA at Unit 2. * * *) However, the margins
to the acceptance criteria of 2.5 Rem at the offsite locations and 5
Rem in the control room are still significant.
* * * * * * *
For the EAB, LPZ, and control room, the post-REA TEDEs are seen
to increase from the values equivalent to the radiation doses from
the current license basis analyses, as they did for the design basis
LRA. (This is attributed to a number of reasons. These include
increase in the fraction of gap activity released to containment,
inclusion of limiting radial peaking in the source term, and
inclusion of alkali metals.) However, the margins to the acceptance
criteria of 6.3 Rem at the offsite locations and 5 Rem in the
control room are still significant * * *.
The changes proposed to the TS for Containment Penetrations are
editorial in nature and will have no effect upon accident
consequences.
The changes proposed to the VFTP TS for the AVS, ABFVES, and
FHVES will not result in a significant increase in any accident
consequences. The changes to make the penetration values for Unit 2
consistent with Unit 1 for the AVS, ABFVES, and FHVES are acceptable
because the appropriate safety factors as delineated in the
applicable regulatory guideline documents are still maintained. The
change to the flowrate specified for the ABFVES is consistent with
the design basis operation of this system. Also, the editorial
changes proposed to the VFTP TS will have no impact on any
accidents.
Operation of the facility in accordance with the proposed
amendment does not involve a significant increase in the
consequences of an accident previously evaluated.
Second Standard
Does operation of the facility in accordance with the proposed
amendment create the possibility of a new or different kind of
accident from any accident previously evaluated? No.
This proposed amendment does not involve addition, removal, or
modification of any plant system, structure, or component. These
changes will not affect the operation of any plant system,
structure, or components as directed in plant procedures.
The analysis performed in support of this license amendment
request, together with the analyses of the design basis fuel
handling accident and weir gate drop reported in previously
submitted and NRC approved license amendment requests, includes full
scope implementation of AST methodology. This analysis does not
represent any change in the post-accident operation of any plant
system, structure, or component.
Operation of the facility in accordance with this amendment does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
Third Standard
Does operation of the facility in accordance with the proposed
amendment involve a significant reduction in the margin of safety?
No.
Margin of safety is related to confidence in the ability of
fission product barriers to perform their design functions following
any of their design basis accidents. These barriers include the fuel
cladding, the Reactor Coolant System, and the containment. The
performance of these barriers either during normal plant operations
or following an accident will not be affected by the changes
associated with the license amendment request.
The AVS is associated with the containment fission product
barrier. Its post-accident operation will not be affected by
implementation of the amendment to its TS. The operation of the
ABFVES either during normal plant operations or following an
accident will not be affected by implementation of the amendment to
its TS. The operation of the FHVES either during normal plant
operations or following an accident will not be affected by
implementation of the amendment to its TS. The operation of the
CRAVS either during normal plant operations or following an accident
will not be adversely affected by the proposed changes to its TS
Bases. The operation of Containment Penetrations following an
accident will not be adversely affected by the proposed change to
its TS.
As noted, an analysis of radiological consequences of the design
basis LOCA at Catawba Nuclear Station has been performed in support
of this license amendment request. The design basis LOCA scenarios
were selected based on extensive evaluations of Catawba, its design
basis, and its anticipated response to a design basis LOCA. Credit
was taken only for safety related systems, structures, and
components in simulating the mitigation of radiological consequences
of the LOCA. Limiting values were taken for performance
characteristics of the Class 1E systems modeled in the analysis. The
radiological consequences (TEDE radiation doses at the EAB, LPZ, and
in the control room) are within the regulatory guideline values with
significant margin.
The changes proposed to the VFTP TS for the AVS, ABFVES, and
FHVES will not result in a significant reduction in the margin of
safety. These changes are supported by regulatory guidance
documents, and are consistent with existing system operation. Also,
the editorial changes proposed to the VFTP TS will not have any
impact on safety.
Operation of the facility in accordance with the proposed
amendment 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: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: September 30, 2004, as supplemented by
letter dated April 26, 2005.
Description of amendment request: The proposed amendment would
change the existing steam generator (SG) tube surveillance program to
be consistent with that being proposed by the Technical Specifications
Task Force (TSTF) in TSTF-449. These proposed changes would revise
Technical Specification (TS) 1.1 on definitions, TS 3.4.13 on reactor
coolant system
[[Page 29791]]
operational leakage, TS 5.5.9 on SG program, and TS 5.6.7 on SG tube
inspection reports, and add a new TS 3.4.16 on SG tube integrity. Also,
as a result of the licensee replacing the SGs with SGs having a new
Alloy 690 thermally treated tubing design, the TSs would be revised to
reflect this replacement. The September 30, 2004, application was
noticed in the Federal Register on November 9, 2004 (69 FR 64987).
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 design basis operating conditions (including startup,
power operation, hot standby, cooldown, anticipated transients and
postulated accidents). The SG performance criteria are based on tube
structural integrity, accident induced leakage, and operational
LEAKAGE. These criteria assure that the probability of an accident
will not be increased.
The primary to secondary accident induced leakage rate for any
design basis accidents, other than an 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. [The primary to secondary accident induced leakage rate is
relatively inconsequential for the SG tube rupture analysis.] The
operational LEAKAGE performance criterion meets current NRC
regulations and NEI [Nuclear Energy Institute] 97-06 criteria for
reactor coolant system (RCS) operational primary to secondary
LEAKAGE through any one SG of 150 gallons per day. These criteria
assure that accident doses will stay within regulatory and licensing
basis limits.
Therefore, the proposed change does not affect the probability
or consequences of any ANO-1 [Arkansas Nuclear One, Unit 1] analyzed
accidents.
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 technical
specifications. 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 proposed change does not affect the design of the
SGs, their method of operation, or primary or secondary coolant
chemistry controls. The proposed 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.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
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 technical specifications.
Therefore, the margin of safety is not changed by the proposed
change to the ANO-1 TSs.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Allen G. Howe.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: March 30, 2005.
Description of amendment request: The proposed amendment adopts the
following Nuclear Regulatory Commission (NRC) approved Technical
Specification Task Force (TSTF) changes that affect the Boiling Water
Reactor (BWR)/6 Improved Standard Technical Specifications:
--------------------------------------------------------------------------------------------------------------------------------------------------------
TSTF No. Description TS section affected Type of change
--------------------------------------------------------------------------------------------------------------------------------------------------------
046, Rev. 1..................... Clarify the Containment Isolation Valve SR 3.6.1.3.4.................... Administrative.
Surveillance Requirement (SR) to apply only to SR 3.6.4.2.2....................
automatic isolation valves. SR 3.6.5.3.3....................
222, Rev. 1..................... Control Rod Scram Time Testing................... SR 3.1.4.1...................... Administrative.
SR 3.1.4.4......................
264, Rev........................ Delete flux monitors specific overlap SRs........ SR 3.3.1.1.5.................... Less Restrictive.
SR 3.3.1.1.6....................
Table 3.3.1.1-1.................
275, Rev. 0..................... Clarify requirements for Diesel Generator (DG) Table 3.3.5.1-1, Footnote (a)... Administrative.
start signal on Reactor Pressure Vessel (RPV)
level--low, low, low during RPV cavity flood-up.
276, Rev. 2..................... Revise DG full load rejection test............... SR 3.8.1.9...................... Less Restrictive.
SR 3.8.1.10.....................
SR 3.8.1.14.....................
300, Rev. 0..................... Eliminate DG loss of coolant accident-Start SRs SR 3.8.2.1...................... Less Restrictive.
while in shutdown when emergency core cooling
system is not required.
322, Rev. 2..................... Secondary Containment Integrity SRs.............. SR 3.6.4.1.3.................... Administrative.
SR 3.6.4.1.4....................
400, Rev. 1..................... Clarification of SR on bypass of DG automatic SR 3.8.1.13..................... Administrative.
trips.
416, Rev. 0..................... SR 3.5.1.2 Notation.............................. LCO 3.5.1....................... Administrative.
SR 3.5.1.2......................
LCO 3.5.2.......................
SR 3.5.2.4......................
--------------------------------------------------------------------------------------------------------------------------------------------------------
[[Page 29792]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the TS [Technical Specifications]
involve both administrative and less restrictive changes. The
administrative changes involve wording changes that clarify
requirements without changing the original intent. As such, these
types of changes do not affect initiators of analyzed events and do
not affect the mitigation of any accidents or transients.
The less restrictive changes involve modifications to
Surveillance Requirements. The modified Surveillance Requirements do
not cause the plant to be operated in a new or different manner and
the required equipment continues to be tested in a manner and at a
frequency necessary to provide confidence that the equipment can
perform its intended safety function. Consequently, no initiators to
accidents previously evaluated are affected and no mitigating
equipment assumed in accidents previously evaluated is adversely
affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not involve a physical alteration of the
plant (no new or different type of equipment will be installed), do
not change the design function of any equipment, and do not change
the methods of normal plant operation. Accordingly, the proposed
changes do not create any new credible failure mechanisms,
malfunctions, or accident initiators not previously considered in
the GGNS [Grand Gulf Nuclear Station] design and licensing basis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a margin
of safety?
Response: No.
The proposed changes have no affect on any safety analysis
assumptions or methods of performing safety analyses. The changes do
not adversely affect system OPERABILITY or design requirements and
the equipment continues to be tested in a manner and at a frequency
necessary to provide confidence that the equipment can perform its
intended safety functions. 10 CFR 50.36 (c)(3) requires the TS to
include Surveillance Requirements relating to test, calibration, or
inspection to assure that the necessary quality of systems and
components is maintained, that facility operation will be within
safety limits, and that the limiting conditions for operation will
be met. The GGNS TS Surveillance Requirements will continue to
provide this assurance with the proposed adoption of the NRC
approved TSTF changes.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
NRC Section Chief: Allen G. Howe.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: December 14, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.3.G, ``Scram Discharge Volume
[SDV],'' to allow vent or drain lines with inoperable valves to be
isolated instead of requiring the valves to be restored to Operable
status or to be in Hot Shutdown within 12 hours.
The NRC staff issued a Notice of Opportunity for Comment in the
Federal Register on February 24, 2003 (68 FR 8637), on possible
amendments to revise the action for one or more SDV vent or drain lines
with an inoperable valve, 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 license
amendment applications in the Federal Register on April 15, 2003 (68 FR
18294). The licensee affirmed the applicability of the model NSHC
determination (modified slightly as a result of the Pilgrim Nuclear
Power Station TS format) in its application dated December 14, 2004.
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.
A change is proposed to allow the affected SDV vent and drain
line to be isolated when there are one or more SDV vent or drain
lines with vent or drain valves inoperable instead of requiring the
valves to be restored to operable status or be in Hot Shutdown
within 12 hours. With one SDV vent or drain valve inoperable in one
or more lines, the isolation function would be maintained since the
redundant valve in the affected line would perform its safety
function of isolating the SDV. Following the completion of the
required action, the isolation function is fulfilled since the
associated line is isolated. The ability to vent and drain the SDV
is maintained and controlled through administrative controls. This
requirement assures the reactor protection system is not adversely
affected by the inoperable valves. With the safety functions of the
valves being maintained, the probability or consequences of an
accident previously evaluated are not significantly increased.
Criterion 2: The proposed change does not 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)
or a change in the methods governing normal plant operation. Thus,
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 proposed change ensures that the safety functions of the SDV
vent and drain valves are fulfilled. The isolation function is
maintained by redundant valves and by the required action to isolate
the affected line. The ability to vent and drain the SDV is
maintained through administrative controls. In addition, the reactor
protection system will prevent filling of the SDV to the point that
it has insufficient volume to accept a full scram. Maintaining the
safety functions related to isolation of the SDV and insertion of
control rods ensures that the proposed change does not involve a
significant reduction in the margin of safety.
Based on the reasoning presented above, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Attorney for licensee: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts 02360-5599.
NRC Section Chief: Darrell J. Roberts.
[[Page 29793]]
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: January 21, 2005.
Description of amendment request: The proposed change permanently
revises Isolation Condenser (IC) Technical Specifications (TS) Section
3.5.3, ``IC System.'' Specifically, surveillance requirement SR 3.5.3.4
is modified by the addition of a note which states the IC System heat
removal capability surveillance is not required to be performed until
12 hours after adequate reactor power is achieved to perform the test.
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:
According to 10 CFR 50.92, ``Issuance of amendment,'' paragraph
(c), a proposed amendment to an operating license involves a no
significant hazards consideration if 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;
(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.
In support of this determination, an evaluation of each of the
three criteria set forth in 10 CFR 50.92 is provided below regarding
the proposed license amendment.
(1) Does the proposed amendment involve a significant increase
in the probability or consequences of an accident previously
evaluated?
Response: No.
The design function of the Isolation Condenser (IC) System is to
provide reactor core cooling in the event that the reactor becomes
isolated from the turbine and the main condenser by closure of the
main steam isolation valves (MSIVs). Although the system is an
Engineered Safety Feature System, no credit for IC System operation
is taken in the accident analysis. The IC System is designed and
installed to provide adequate core cooling, thereby mitigating the
consequences of this reactor isolation transient (e. g., inadvertent
closure of the MSIVs). This transient has been evaluated in the
Updated Final Safety Analysis Report (UFSAR) as an event of moderate
frequency. The IC system is designed to operate automatically or
manually to perform its design function for reactor pressures
greater than 150 psig. Since the IC System is not credited, this TS
change does not impact any of the assumptions, inputs, or results of
the UFSAR reactor isolation analysis.
The addition of the note to the Technical Specifications
surveillance requirement does not alter the IC System design
function or the processes and parameters by which the system and its
components perform its function. The addition of this note allows
the plant to enter an operating mode necessary to allow performance
of the heat removal capability surveillance. The purpose of this
heat removal capability surveillance is to verify proper flow path
and the ability to remove a design heat load. The proposed change
does not alter the ability or methods used to verify flow path or
heat removal capability. Nor does the change alter the acceptance
criteria for satisfactory performance. Therefore, the change does
not result in an increase in the consequences of a reactor isolation
transient. Additionally, there are no IC System malfunctions or
component failures that could initiate a reactor isolation
transient. The proposed change does not alter the system or its
operation and will not change the IC System's impact on initiating
accidents or transients. Therefore, this change, and any associated
impacts, will not increase the probability of the occurrence of an
accident or transient.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(2) Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The addition of the note to the Technical Specifications
surveillance requirement does not alter the IC System design
function or the processes and parameters by which the system and its
components perform its function. The existing Technical
Specification does not provide any limitations on when the IC System
heat removal capability surveillance may be performed. Present plant
procedures perform this surveillance at between 60% and 75% reactor
power to ensure sufficient steam is available to simulate design
heat loads. The addition of the note to the Technical Specification
does not create any constraints on plant operating conditions
associated with performance of the IC System heat removal capability
surveillance. Operation of the IC System to perform the required
surveillance in operating Modes 1, 2, or 3 has been previously
evaluated and is presently allowed.
The proposed change does not modify the procedural steps for
performing the Technical Specification required surveillance. Nor
does the change alter the methodology for evaluating acceptable
performance. No physical or operational changes are made that could
result in plant or system operation in conditions not previously
evaluated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Technical Specification surveillance requirement SR 3.5.3.4
requires verification of the IC System's heat removal capability
every 60 months. This surveillance ensures the proper system flow
path and ability to remove decay heat following a reactor isolation.
The methodology and acceptance criteria for this surveillance are
not impacted by this change. Technical Specifications presently
allow performance of this surveillance in Modes 1, 2, or 3 and plant
procedures presently perform this surveillance in Mode 1. The
surveillance is still required to demonstrate the IC System design
basis capability of removing the design requirement of 252.5 x
106 Btu/hr. Other IC System surveillance requirements are
not directly or indirectly impacted by this change. Additionally,
this amendment request results in no change to the system's
actuation response, operation, or setpoints for performance.
Therefore, the proposed 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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Gene Y. Suh.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: June 11, 2004.
Brief description of amendment request: The proposed license
amendment request would relocate surveillance test intervals of various
Technical Specification (TS) surveillance requirements to a new program
controlled in accordance with the requirements of 10 CFR 50.59. The
proposed changes would add a new program, the Surveillance Frequency
Control Program, to the Administrative Controls section of the TSs. The
proposed amendment is a pilot submittal in support of the Boiling Water
Reactor Owners' Group Risk-Informed Initiative 5b, ``Relocate
Surveillance Test Intervals to Licensee Control.''
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
[[Page 29794]]
consequences of an accident previously evaluated?
Response: No. The proposed change involves the relocation of
various surveillance test intervals from Technical Specifications
(TS) to a licensee-controlled program and is administrative in
nature. The proposed change does not involve the modification of any
plant equipment or affect basic plant operation. The proposed change
will have no impact on any safety related structures, systems or
components. Surveillance test intervals are not assumed to be an
initiator of any analyzed event, nor are they assumed in the
mitigation of consequences of accidents. The surveillance
requirements themselves will be maintained in TS[s] along with the
applicable Limiting Conditions for Operation (LCOs) and Action
statements. The surveillances performed at the intervals specified
in the licensee-controlled program will assure that the affected
system or component function is maintained, that the facility
operation is within the Safety Limits, and that the LCOs are met.
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 previously evaluated?
Response: No. The proposed change does not involve any physical
alteration of plant equipment and does not change the method by
which any safety-related system performs its function or is tested.
As such, no new or different types of equipment will be installed,
and the basic operation of installed equipment is unchanged. The
methods governing plant operation and testing remain consistent with
the safety analysis assumptions.
Therefore, the proposed change will 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 is administrative in nature,
does not negate any existing requirement, and does not adversely
affect existing plant safety margins or the reliability of the
equipment assumed to operate in the safety analysis. As such, there
are no changes being made to safety analysis assumptions, safety
limits or safety system settings that would adversely affect plant
safety as a result of the proposed change. Margins of safety are
unaffected by relocation of the surveillance test intervals to a
licensee-controlled program.
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: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Darrell J. Roberts.
Exelon Generation Company, LLC, Docket No. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: July 22, 2004, as supplemented December
3, 2004.
Description of amendment request: The proposed amendment would
modify the operability and surveillance requirements in Technical
Specification 3/4.1.3, ``Control Rods.'' Specifically, the proposed
changes would (1) exclude a fully inserted immovable control rod from
the shutdown action statement, (2) eliminate consideration of control
rod drive water pressure in the action statement, and (3) limit the 24-
hour exercise test of other control rods to a one-time occasion
following detection of an immovable control rod.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No. The first proposed change would exclude fully
inserted immovable control rods from consideration in the plant
shutdown action statement. An inoperable control rod that has been
fully inserted, and disarmed, has satisfied the safety function of
that control rod since it is in a position of maximum contribution
to shutdown capability. A plant shutdown for this situation would
result in an unnecessary plant thermal cycle without any
compensatory safety benefit. Under the proposed change, inoperable
inserted rods would continue to be counted in the operability
requirement precluding power operation with more than 8 inoperable
control rods.
The second proposed change removes the control rod drive (CRD)
water pressure limits from the insertion capability test of
inoperable, non-stuck, control rods. Reactor pressure, assisted by a
pre-charged accumulator, provides the driving force for the rapid
shutdown of the reactor (scram), independent of the CRD water
pressure. Variation of this pressure is not an indicator of a
degraded control rod, and does not inhibit the safety function of
the control rod. Control rod scram and exercise testing requirements
assure the operability of the CRD system. The proposed change would
eliminate the need to unnecessarily insert a control rod into the
core if it could not be repositioned using the normal drive water
pressure setting.
The third proposed change would limit the increased frequency
surveillance requirement (every 24 hours) exercise test of withdrawn
control rods upon discovery of an immovable control rod to a one-
time test in lieu of every 24 hours. A one-time 24-hour test is
sufficient to determine if a generic control rod problem exists.
Under the proposed change, following the 24-hour test, and in
absence of any additional detectable problems, the control rod
exercise test would revert back to a normal testing frequency.
Repetitive 24-hour tests [have] the potential to reduce the operable
lifespan of hydraulic control unit components and increases the
potential for a reactivity management event.
The proposed changes will not impede the ability of the
surveillance requirements to detect control rod degradation, or
inhibit the control rod drive system from performing its designed
safety function.
Therefore, this proposed amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any previously evaluated?
Response: No. The proposed changes do not alter the physical
design, safety limits, or safety analysis assumptions, associated
with the operation of the plant. Accordingly, the changes do not
introduce any new accident initiators, nor do they reduce or
adversely affect the capabilities of any plant structure, system, or
component to perform their safety function.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No. A fully inserted [control] rod has satisfied its
safety function by being in the position of maximum contribution to
shutdown reactivity. Eliminating the CRD water pressure limits does
not impact scram capability. Further, the proposed changes will
eliminate extended accelerated control rod testing that may shorten
the lifespan of control components without any compromise in the
detection of control rod operability problems. The proposed changes
would not impact control rod operability and surveillance
requirements that are necessary to assure that the control rod
system will perform its designed safety function.
Therefore, the proposed amendment 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: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
[[Page 29795]]
NRC Section Chief: Darrell J. Roberts.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: April 20, 2005.
Description of amendment request: The proposed amendment would
revise the technical specifications (TSs) to replace plant-specific
position titles with generic position titles. The proposed changes are
consistent with NUREG-1430, ``Standard Technical Specifications--
Babcock and Wilcox Plants,'' Revision 3. Also, the licensee proposes to
delete TS 6.7, ``Safety Limit Violation or Protective Limit
Violation,'' including a change to TS 2.1.2, ``Safety Limits and
Limiting Safety System Settings--Reactor Core,'' associated with the
deletion of TS 6.7. Additionally, the licensee proposes to relocate to
the Technical Requirements Manual (TRM), the Process Control Program
requirements from TS 6.8, ``Procedures and Programs,'' and from TS
6.14, ``Process Control Program (PCP).'' Associated with this change,
TS Definition 1.30, ``Process Control Program,'' is proposed to be
deleted. Also, TS 6.15, ``Offsite Dose Calculation Manual (ODCM),'' is
proposed to be modified to eliminate the requirement that changes to
the ODCM be reviewed and accepted by the Plant Operations Review
Committee (PORC). Lastly, the licensee proposes to revise in the TS the
title, ``Industrial Security Plan'' to ``Physical Security Plan.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes affect the requirements for the
administrative controls section of the Technical Specifications. The
proposed changes are primarily intended to make the plant-specific
position/organizational titles found in the administrative controls
section of the Technical Specifications more generic. The proposed
changes do not affect any plant structures, systems, and components,
and have no effect on plant operations. The proposed changes are
administrative and do not affect any existing limits. Accident
initial conditions, probability, and assumptions remain as
previously analyzed. The proposed changes will have no effect on
accident initiation frequency. The proposed changes do not
invalidate the assumptions used in evaluating the radiological
consequences of any accident. Therefore, the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes are administrative and do not introduce any
new or different accident initiators. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes are administrative and will not have a
significant effect on any margin of safety. Therefore, the proposed
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Gene Y. Suh.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: April 22, 2005.
Description of amendment request: The proposed amendment would
revise the technical specifications (TSs) related to fuel handling and
storage. Specifically, the proposed change is to reflect that spent
fuel storage racks are no longer installed in the cask pit or transfer
pit and that there are no longer any low-density fuel storage racks in
the spent fuel pool. Additionally, the proposed changes would relocate
the requirements of TS 3/4.9.7, ``Crane Travel--Fuel Handling
Building,'' to the Technical Requirements Manual (TRM).
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 amendment would relocate the requirements of TS 3/
4.9.7 to the DBNPS [Davis-Besse Nuclear Power Station] TRM. Any
subsequent changes to the TRM would require evaluation under the
appropriate regulatory processes (e.g., 10 CFR 50.59). The proposed
relocation of TS 3/4.9.7 does not affect any accident initiators.
The relocated TRM requirements will assure the initial conditions
assumed in the analysis of a fuel handling accident are maintained.
The proposed change does not affect the ability of plant equipment
to mitigate the consequences of any accident. The proposed changes
to reflect that fuel storage racks are no longer installed in the
cask pit or transfer pit and that low density fuel storage racks are
no longer installed in the spent fuel pool are consistent with the
current plant configuration. The proposed changes do not affect any
accident initiators. The revised requirements will continue to
assure the capability to mitigate the consequences of a fuel
handling accident in the fuel storage area. Therefore, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed relocation of TS 3/4.9.7 to the TRM does not alter
the design, operation, or testing of any structure, system, or
component. The proposed changes to reflect that fuel storage racks
are no longer installed in the cask pit or transfer pit and that low
density fuel storage racks are no longer installed in the spent fuel
pool are consistent with the current plant configuration. No new
accident initiators are 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 relocation of TS 3/4.9.7 to the TRM does not alter
the design, operation, or testing of any structure, system, or
component. The proposed changes to reflect that fuel storage racks
are no longer installed in the cask pit or transfer pit and that low
density fuel storage racks are no longer installed in the spent fuel
pool are consistent with the current plant configuration and do not
adversely affect the ability of any structure, system, or component
to perform its safety function. 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: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
[[Page 29796]]
NRC Section Chief: Gene Y. Suh.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: May 2, 2005.
Description of amendment request: The proposed amendment would
revise technical specification (TS) Figure 2.1-1, ``Reactor Core Safety
Limit'' and TS Table 2.2-1, ``Reactor Protection System Instrumentation
Trip Setpoints.'' These TS revisions would support the use of Framatome
Mark B-HTP fuel in the reactor.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes include a revision of the Reactor Core
Safety Limits specified in Technical Specification (TS) Section
2.1.1, and a revision of the Reactor Protection System (RPS) Reactor
Coolant System (RCS) Pressure-Temperature setpoint Allowable Value
provided in TS Section 2.2.1. The proposed changes preserve the
design DNB [departure from nucleate boiling] Ratio safety criterion
that there shall be at least a 95% probability at a 95% confidence
level that the hot fuel rod in the core does not experience a
departure from nucleate boiling during normal operation or events of
moderate frequency. Further, there are no evaluated accidents in
which the fuel cladding or fuel assembly structural components are
assumed to arbitrarily fail as an accident initiator. The fuel
handling accident analysis assumes that the cladding does, in fact,
fail as a result of an undefined fuel handling event. However, the
probability of an accident initiator for the fuel handling accident
is independent of the parameters changed in this amendment request.
In addition, the proposed changes do not involve a significant
increase in the consequences of an accident previously evaluated
because the proposed changes do not alter any assumptions previously
made in the radiological consequence evaluations, or affect
mitigation of the radiological consequences of an accident
previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
because no new accident scenarios, failure mechanisms or single
failures are introduced as a result of the proposed. All systems,
structures, and components previously required for the mitigation of
an event remain capable of fulfilling their intended design
function.
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 proposed changes do not involve a significant reduction in a
margin of safety because extensive analyses of the primary fission
product barriers, conducted in support of the proposed changes, have
concluded that all relevant design criteria remain satisfied, both
from the standpoint of the integrity of the primary fission product
barrier and from the standpoint of compliance with the regulatory
acceptance criteria. As appropriate, all evaluations have been
performed using methods that have either been reviewed and approved
by the Nuclear Regulatory Commission or that are in compliance with
applicable regulatory review guidance and standards.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Gene Y. Suh.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: January 10, 2005.
Description of amendment request: The amendment request proposes to
revise the surveillance interval associated with Technical
Specification Surveillance Requirement 4.6.1.3b from once every 6
months to once every 24 months for verification that only one door in
each containment air lock can be opened at a time.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed amendment will neither effect nor change any design
function, or method of performing or controlling design functions,
or any analysis that verifies the capability of structures, systems
and components (SSCs) to perform their designed function(s). The
proposed amendment will have no adverse effect on plant operation or
its controlled configuration. As a result, the proposed amendment
will not change assumptions, or change, degrade or prevent actions
described or assumed in accidents evaluated and described in the
Seabrook Station Updated Final Safety Analysis Report (UFSAR). The
proposed change extends the surveillance interval from 6 months to
24 months to verify proper functioning of the containment air lock
interlocks. The proposed change to the Surveillance Requirement
testing interval does not adversely affect performance of the
Surveillance Requirement that verifies the functional status of the
air lock interlock to prevent both air lock doors to be open
simultaneously. Containment integrity is not affected by the
proposed amendment. The radiological consequences of an event are
unchanged, since the functional status of the air lock interlock is
not adversely affected and the air lock doors' ability to withstand
the maximum expected post accident containment pressure is not
adversely affected by the proposed change. Therefore, the proposed
amendment does not adversely affect nuclear safety or continued safe
operation of Seabrook Station, or result in an increase in the
radiological consequences of any accident described in the Seabrook
Station UFSAR.
Therefore, it is concluded that the proposed change does not
involve a significant increase in the probability or consequence of
an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed amendment will neither effect nor change any design
function, or method of performing or controlling design functions,
or any analysis that verifies the capability of structures, systems
and components (SSCs) to perform their designed function(s). The
proposed amendment will have no adverse effect on plant operation or
its controlled configuration. As a result, the proposed amendment
will not change assumptions, or change, degrade or prevent actions
described or assumed in accidents evaluated and described in the
Seabrook Station UFSAR. There are no changes associated with
extending the surveillance interval for the air lock interlock that
could potentially introduce new failure modes or accident
initiators.
Therefore, it is concluded that the proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The proposed change extends the surveillance interval from 6
months to 24
[[Page 29797]]
months to verify proper functioning of the containment air lock
interlock. The containment air lock interlocks are normally not
challenged and operating experience has shown these components have
an excellent surveillance pass rate. Furthermore, increasing the
surveillance interval has no affect on the air lock doors' ability
to withstand the maximum expected post accident containment
pressure. Containment integrity is not affected by the proposed
amendment. The proposed amendment will neither effect nor change any
design function, or method of performing or controlling design
functions, or any analysis that verifies the capability of
structures, systems and components (SSCs) to perform their designed
function(s). The functional status of the containment air lock
interlocks will continue to be verified.
Therefore, it is concluded that the proposed 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: M. S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Section Chief: Darrell J. Roberts.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: March 28, 2005.
Description of amendment request: The proposed amendment would
extend the expiration date of Facility Operating License (FOL) NPF-86
for Seabrook Station, Unit No. 1 by approximately 3.4 years. The
extension would set the date of expiration of the FOL to occur 40 years
from the date of issuance of the full-power operating license.
Specifically, the FOL, with a current expiration date of October 17,
2026 would be revised to expire on March 15, 2030. This change would
allow the recapture of zero-power and low-power testing time in
accordance with SECY-98-296, ``Agency Policy Regarding Licensee
Recapture of Low-Power Testing or Shutdown Time for Nuclear Power
Plants,'' dated December 21, 1998.
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?
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated since it does not involve a change to design configuration
or operation of the facility. The proposed change does not effect
the source term, containment isolation or radiological release
assumptions used in evaluating the radiological consequences of an
accident previously evaluated in the Seabrook Station UFSAR [updated
final safety analysis report]. In addition, Seabrook Station Unit
[No.] 1 was designed and constructed to ensure a 40-year service
life. Design features were incorporated that provide for inspection
of structures, systems and components during the 40-year service
life. Surveillance, inspection and maintenance practices have been
implemented in accordance with the American Society of Mechanical
Engineers Boiler and Pressure Vessel Code and the unit Technical
Specifications to provide assurance that any degradation in plant
safety-related equipment will be identified and corrected to provide
continued safe operation of the unit throughout the duration of the
facility operating license.
The recapture period requested by this amendment is for 3.4
years. This time is insignificant from an aging effect perspective
when considered in conjunction with the surveillance, inspection and
maintenance programs implemented to provide early indication of
degradation in plant safety-related equipment. Continual maintenance
and testing provides for continued safe operation of the unit
throughout the duration of the facility operating license.
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?
The proposed amendment revises the expiration of the facility
operating license such that the expiration of the facility operating
license is based upon issuance of the FPOL [full-power operating
license] and not upon issuance of the ZPOL/LPOL [zero-power
operating license/low-power operating license]. The proposed
change[s] do[es] not involve physical alteration of plant systems[,]
structures or components or changes in parameters governing the
manner in which the plant is operated and maintained.
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?
The proposed amendment revises the expiration of the facility
operating license such that the expiration of the facility operating
license is based upon issuance of the FPOL and not upon issuance of
the ZPOL/LPOL. No physical changes are being made to the design
features or operation of the facility.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel cladding, reactor coolant
system pressure boundary and the containment structure) to limit the
radiological dose to the public and control room operators in the
event of an accident. The proposed amendment to the facility
operating license has no impact on the margin of safety and
robustness provided in the design and construction of the facility.
In addition, the proposed amendment will not relax any of the
criteria used to establish safety limits, nor will the proposed
amendment relax safety system settings or limiting conditions of
operation as defined in the Technical Specifications.
Therefore, the proposed amendment does not result in 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: M.S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Section Chief: Darrell J. Roberts.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of amendment request: April 26, 2005.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TS) 5.6.5.b., ``Core Operating Limits
Report (COLR),'' to add the Palisades-specific fuel assembly growth
model to the analytical methods referenced in the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed license amendment augments an existing analytical
method used to determine the core operating limits per Technical
Specification 5.6.5.b. Accidents previously evaluated will be
unaffected because they will continue to be analyzed using
applicable methodologies approved by the Nuclear Regulatory
Commission to ensure all required safety limits are met. The
proposed amendment does not affect the acceptance criteria for any
Final Safety Analysis Report (FSAR) safety analysis analyzed
accidents and anticipated operational occurrences. As such, the
proposed amendment does not increase the probability or consequences
of an accident. The proposed amendment does not involve
[[Page 29798]]
operation of the required structures, systems or components (SSCs)
in a manner or configuration different from those previously
recognized or evaluated.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not involve a physical alteration of
any SSC or a change in the way any SSC is operated. The proposed
amendment does not involve operation of any required SSCs in a
manner or configuration different from those previously recognized
or evaluated. No new failure mechanisms will be introduced by the
changes being requested.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed amendment does not, by itself, introduce a failure
mechanism. The proposed amendment does not involve any physical
changes to the plant or manner in which the plant is operated. The
proposed changes do not affect the acceptance criteria for any FSAR
safety analysis analyzed accidents or anticipated operational
occurrences. All required safety limits would continue to be
analyzed using methodologies approved by the Nuclear Regulatory
Commission.
Therefore, the proposed amendment would 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: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Section Chief: L. Raghavan.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: November 23, 2004.
Description of amendment request: The proposed amendment revises
the descriptive wording of Technical Specifications Table 1-1, ``RPS
[reactor protection system] Limiting Safety System Settings,'' for the
Reactor Trip setpoint for Low Steam Generator Water Level to relocate
unnecessary detail and converts Technical Specifications Section 4.0,
Design Features, to be consistent with NUREG-1432, Revision 3,
``Standard Technical Specifications for Combustion Engineering
Plants.'' These changes will be needed to support the operation of Fort
Calhoun Station (FCS) after major components (steam generators,
pressurizer, and reactor vessel head) are replaced in 2006.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes are not related to an initiator of any
previously evaluated accident. The proposed changes revise
descriptive information only, and 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. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes only relocate descriptive information in
the Technical Specifications to the USAR [Updated Safety Analysis
Report]. Modifications will not be made to existing equipment nor
will any new or different types of equipment be installed. The
proposed changes to the Technical Specifications will not alter
assumptions made in safety analysis and licensing bases.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed administrative changes only relocate descriptive
information in the FCS Technical Specifications to the USAR, and
have no effect on safety margins.
Therefore, this technical specification 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,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: September 8, 2004.
Description of amendment request: The proposed amendments would
change the SSES 1 and 2 Technical Specifications (TSs) limiting
conditions for operation (LCO) 3.8.4, ``DC Sources-Operating,'' to
incorporate the Technical Specifications Change Task Force (TSTF) 16,
Revision 2, and other unrelated editorial changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability of occurrence [sic] or consequences of an accident
previously evaluated?
Response: No.
The Technical Specification allowed Completion Time for any
inoperability is not an initiator to any accident sequence analyzed
in the Final Safety Analysis Report (FSAR). The changes do not
involve any physical change to structures, systems, or components
(SSCs) and does not alter the method of operation or control of
SSCs. The current assumptions in the safety analysis regarding
accident initiators and mitigation of accidents are unaffected by
these changes. No additional failure modes or mechanisms are being
introduced and the likelihood of previously analyzed failures
remains unchanged.
Operation in accordance with the proposed Technical
Specification (TS) ensures that the AC distribution system and
supported equipment functions remain capable of performing the
function as described in the FSAR. Therefore, the mitigative
functions supported by the system will continue to provide the
protection assumed by the analysis.
The correction of typographical errors, changes in format and
the deletion of a no longer required one-time exemption are
administrative changes.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. There
are no setpoints, at which protective or mitigative actions are
initiated, affected by this change. This change will not alter the
manner in which equipment operation is initiated, nor will the
[[Page 29799]]
function demands on credited equipment be changed. No alterations in
the procedures that ensure the plant remains within analyzed limits
are being proposed, and no changes are being made to the procedures
relied upon to respond to an off-normal event as described in the
FSAR. The correction of typographical errors, changes in format and
the deletion of a no longer required one-time exemption are
administrative changes. As such, no new failure modes are being
introduced. The change does not alter assumptions made in the safety
analysis and licensing basis.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed change is acceptable because the
restoration times for deenergized AC distribution subsystems has
been previously evaluated in Unit 2 Amendment No. 148. Additional
margin of safety is gained with the inclusion of the requirement to
enter applicable actions for inoperable Class lE battery chargers as
a result of inoperable AC bus(es). The correction of typographical
errors, changes in format and the deletion of a no longer required
one-time exemption are administrative changes. Therefore the plant
response to analyzed events will continue to provide the margin of
safety assumed by the analysis.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: Richard J. Laufer.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: January 28, 2005.
Description of amendment request: The proposed amendments would
change the SSES 1 and 2 Technical Specifications (TSs) 5.5.6,
``Inservice Testing Program,'' to replace the reference to American
Society of Mechanical Engineers (ASME) Boiler and PressureVessel Code,
Section XI, with a reference to ASME Code for Operation and Maintenance
of Nuclear Power Plants (ASME OM Code) as the source of requirements
for the inservice testing of ASME Code Class 1, 2, and 3 pumps and
valves. These changes are consistent with the implementation of the
SSES 1 and 2 Third 10-Year Interval Inservice Testing Program in
accordance with the requirements of 10 CFR 50.55a(f).
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 of occurrence [sic] or consequences of an accident
previously evaluated?
Response: No.
The proposed changes revise Technical Specification 5.5.6 for
SSES Units 1 and 2 to conform to the requirements of 10 CFR
50.55a(f) regarding the inservice testing of pumps and valves for
the Third 10-Year Interval. The current Technical Specifications
reference the ASME Boiler and Pressure Vessel Code, Section XI,
requirements for the inservice testing of ASME Code Class 1, 2, and
3 pumps and valves. The proposed changes would reference the ASME OM
Code, which is consistent with 10 CFR 50.55a(f) and accepted for use
by the NRC. The proposed changes are administrative in nature.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes revise Technical Specification 5.5.6 for
SSES Units I and 2 to conform to the requirements of 10 CFR
50.55a(f) regarding the inservice testing of pumps and valves for
the Third 10-Year Interval. The current Technical Specifications
reference the ASME Boiler and Pressure Vessel Code, Section XI,
requirements for the inservice testing of ASME Code Class 1, 2, and
3 pumps and valves. The proposed changes would reference the ASME OM
Code, which is consistent with 10 CFR 50.55a(f)and accepted for use
by the NRC. The proposed changes are administrative in nature.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes revise Technical Specification 5.5.6 for
SSES Units I and 2 to conform to the requirements of 10 CFR
50.55a(f) regarding the inservice testing of pumps and valves for
the Third 10-Year Interval. The current Technical Specifications
reference the ASME Boiler and Pressure Vessel Code, Section XI,
requirements for the inservice testing of ASME Code Class 1, 2, and
3 pumps and valves. The proposed changes would reference the ASME OM
Code, which is consistent with 10 CFR 50.55a(f) and accepted for use
by the NRC. The proposed changes are administrative in nature.
Therefore, the proposed change[s] does [sic] not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: Richard J. Laufer.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: February 7, 2005.
Description of amendment request: The proposed amendments would
change the SSES 1 and 2 Technical Specifications (TSs) for ``Secondary
Containment,'' limiting condition for operation (LCO) 3.6.4.1, by
revising the frequency note applicable to Surveillance Requirements
(SR) 3.6.4.1.4 and SR 3.6.4.1.5. The revised note requires each SR be
performed with the 3 zone configuration every 60 months.
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 does not involve a significant increase in
the probability of an accident previously evaluated because neither
Secondary Containment nor the Standby Gas Treatment System is an
initiator of an accident. Both mitigate accident consequences.
The consequences of a Design Basis Analysis-Loss of Coolant
Accident (DBA-LOCA) have been evaluated in the FSAR [final safety
analysis report]. Revising the surveillance frequency to require the
most limiting configurations to be tested with the 60-month period
rather than just the three zone configuration provides assurance
that the most limiting secondary containment configuration is tested
every 60 months in accordance with the original intent of the
surveillance frequency. The proposed change also provides added
assurance of acceptable performance within the analysis assumptions
of the FSAR. The radiological evaluation of DBA-LOCA doses,
including doses offsite, control room habitability, and exposures
for personnel are not impacted.
[[Page 29800]]
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not involve a physical alteration of the
plant. No new or different [kind] of equipment will be installed nor
will there be changes in methods governing normal plant operation.
The potential for the loss of plant systems or equipment to
mitigate the effects of an accident is not altered.
The proposed changes do not require any new operator response or
introduce any new opportunities for operator error not previously
considered.
Thus, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in
[a] margin of safety?
Response: No.
The proposed change does not involve a significant reduction in
[a] margin of safety.
The surveillance test change ensures all the secondary
containment configurations are tested within a 60-month period when
only one configuration was previously required to be tested. This
change has a positive effect on the margin of safety as it provides
more restrictive testing requirement that will provide added
assurance of acceptable secondary containment performance.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: Richard J. Laufer.
PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric
Station, Unit 2 (SSES 2), Luzerne County, Pennsylvania
Date of amendment request: March 18, 2005.
Description of amendment request: The proposed amendment would
revise the SSES 2 Technical Specification (TS) 3.3.8.1, ``Loss of Power
(LOP) Instrumentation,'' to: (1) clarify that Condition A applies to
inoperable instrumentation other than during the performance of
Surveillance Requirement (SR) 3.8.1.19 loss-of-coolant accident/loss of
offsite power testing on Unit 1 and to revise TS Bases section to
clarify that this condition is applicable to both Unit 1 and Unit 2 LOP
Instrumentation, (2) add new Condition B to allow the LOP
instrumentation for two Unit 1 4.16kV Engineered Safeguards System
buses in the same Division to be inoperable for up to 8 hours for the
performance of SR 3.8.1.19 on Unit 1. In addition, the proposed
amendment would revise the SSES 2 TS 3.8.7, ``Distribution Systems-
Operating,'' to: (1) eliminate ``or more'' and the plural to subsystems
such that the condition would read ``One Unit 1 AC [alternating
current] electrical power distribution subsystem inoperable,'' (2) add
new Condition D for two Unit 1 AC electrical power distribution
subsystems inoperable.
This will impose an 8-hour Completion Time for restoration of at
least one of the two Unit 1 AC distribution subsystems.
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 of occurrence or consequences of an accident
previously evaluated?
Response: No.
The Technical Specification allowed Completion Time for any
inoperability is not an initiator to any accident sequence analyzed
in the Final Safety Analysis Report (FSAR). The changes do not
involve any physical change to structures, systems, or components
(SSCs) and does not alter the method of operation or control of
SSCs. The current assumptions in the safety analysis regarding
accident initiators and mitigation of accidents are unaffected by
these changes. No additional failure modes or mechanisms are being
introduced and the likelihood of previously analyzed failures
remains unchanged.
Operation in accordance with the proposed Technical
Specification (TS) ensures that the AC distribution system and
supported equipment functions remain capable of performing the
function as described in the FSAR [final safety analysis report].
Therefore, the mitigative functions supported by the system will
continue to provide the protection assumed by the analysis.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. There
are no setpoints, at which protective or mitigative actions are
initiated, affected by this change. This change will not alter the
manner in which equipment operation is initiated, nor will the
function demands on credited equipment be changed. No alterations in
the procedures that ensure the plant remains within analyzed limits
are being proposed, and no changes are being made to the procedures
relied upon to respond to an off-normal event as described in the
FSAR. As such, no new failure modes are being introduced. The change
does not alter assumptions made in the safety analysis and licensing
basis.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed change is acceptable because the
restoration time for deenergized AC distribution subsystems has been
previously evaluated in Unit 2 Amendment No. 148. Therefore[,] the
plant response to analyzed events will continue to provide the
margin of safety assumed by the analysis.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: Richard J. Laufer.
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: March 4, 2005.
Description of amendment request: The proposed amendment would
change Technical Specification (TS) 3.5.1, ``Accumulators,'' to extend
the completion time (CT) for Action (a) from 1 hour to 24 hours. The
accumulators are part of the emergency core cooling system and consist
of tanks partially filled with borated water and pressurized with
nitrogen gas. The contents of the tank are discharged to the reactor
coolant system (RCS) if, as during a loss-of-coolant accident (LOCA),
the coolant pressure decreases to below the accumulator pressure.
Action (a) of TS 3.5.1 specifies a CT to restore an accumulator to
operable status when it has been declared inoperable for a reason other
than the boron concentration of the water in the accumulator not being
within the required range. This change was proposed by the Westinghouse
Owners Group participants in the TS Task Force (TSTF) and is designated
TSTF-370. TSTF-370 is supported by NRC-
[[Page 29801]]
approved Topical Report WCAP-15049-A, ``Risk-Informed Evaluation of an
Extension to Accumulator Completion Times,'' submitted on May 18, 1999.
The NRC staff issued a Notice of Opportunity for Comment in the Federal
Register on July 15, 2002 (67 FR 46542), on possible amendments
concerning TSTF-370, 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 license
amendment applications in the Federal Register on March 12, 2003 (68 FR
11880). The licensee affirmed the applicability of the following NSHC
determination in its application dated March 4, 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 basis for the accumulator limiting condition for operation
(LCO), as discussed in Bases Section 3.5.1, is to ensure that a
sufficient volume of borated water will be immediately forced into
the core through each of the cold legs in the event the RCS pressure
falls below the pressure of the accumulators, thereby providing the
initial cooling mechanism during large RCS pipe ruptures. As
described in Section 9.2 of the WCAP-15049, ``Risk-Informed
Evaluation of an Extension to Accumulator Completion Times,''
evaluation, the proposed change will allow plant operation with an
inoperable accumulator for up to 24 hours, instead of 1 hour, before
being required to begin shutdown. The impact of the increase in the
accumulator CT on core damage frequency for all the cases evaluated
in WCAP-15049 is within the acceptance limit of 1.0E-06/yr for a
total plant core damage frequency less than 1.0E-03/yr. The
incremental conditional core damage probabilities calculated in
WCAP-15049 for the accumulator CT increase meet the criterion of 5E-
07 in Regulatory Guides (RGs) 1.174 [``An Approach for Using
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis''] and 1.177 [``An Approach
for Plant-Specific, Risk-Informed Decisionmaking: Technical
Specifications''] for all cases except those that are based on
design basis success criteria. As indicated in WCAP-15049, design
basis accumulator success criteria are not considered necessary to
mitigate large-break LOCA events, and were only included in the
WCAP-15049 evaluation as a worst-case data point. In addition, WCAP-
15049 states that the NRC has indicated that an incremental
conditional core damage frequency greater than 5E-07 does not
necessarily mean the change is unacceptable.
The proposed TS change does not involve any hardware changes nor
does it affect the probability of any event initiators. There will
be no change to normal plant operating parameters, engineered safety
feature actuation setpoints, accident mitigation capabilities,
accident analysis assumptions or inputs.
Therefore, this change does not involve a significant increase
in the probability or consequences of an 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
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed change. As described in Section 9.1 of the WCAP-
15049 evaluation, the plant design will not be changed with this
proposed TS CT increase. All safety systems still function in the
same manner and there is no additional reliance on additional
systems or procedures. The proposed accumulator CT increase has a
very small impact on core damage frequency. The WCAP-15049
evaluation demonstrates that the small increase in risk due to
increasing the accumulator allowed outage time is within the
acceptance criteria provided in RGs 1.174 and 1.177. No new
accidents or transients can be introduced with the requested change
and the likelihood of an accident or transient is not impacted.
The malfunction of safety related equipment, assumed to be
operable in the accident analyses, would not be caused as a result
of the proposed TS change. No new failure mode has been created and
no new equipment performance burdens are imposed.
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Criterion 3: The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not involve a significant reduction in
a margin of safety. There will be no change to the departure from
nucleate boiling ratio (DNBR) correlation limit, the design DNBR
limits, or the safety analysis DNBR limits.
The basis for the accumulator LCO, as discussed in Bases Section
3.5.1, is to ensure that a sufficient volume of borated water will
be immediately forced into the core through each of the cold legs in
the event the RCS pressure falls below the pressure of the
accumulators, thereby providing the initial cooling mechanism during
large RCS pipe ruptures. As described in Section 9.2 of the WCAP-
15049 evaluation, the proposed change will allow plant operation
with an inoperable accumulator for up to 24 hours, instead of 1
hour, before being required to begin shutdown. The impact of this on
plant risk was evaluated and found to be very small. That is,
increasing the time the accumulators will be unavailable to respond
to a large LOCA event, assuming accumulators are needed to mitigate
the design basis event, has a very small impact on plant risk. Since
the frequency of a design basis large LOCA (a large LOCA with loss
of offsite power) would be significantly lower than the large LOCA
frequency of the WCAP-15049 evaluation, the impact of increasing the
accumulator CT from 1 hour to 24 hours on plant risk due to a design
basis large LOCA would be significantly less than the plant risk
increase presented in the WCAP-15049 evaluation.
Therefore, this change does not involve a significant reduction in
a margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the NRC staff proposes to
determine that the requested change does not involve a significant
hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: Darrell J. Roberts.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: August 5, 2004, as superceded in its
entirety by letter dated March 15, 2005.
Brief description of amendments: The proposed amendments would
revise Technical Specification (TS) 3.7.10 entitled ``Control Room
Emergency Filtration/Pressurization System (CREFS)'' to extend the
Completion Time for ACTION B., ``Two CREFS Trains inoperable due to
inoperable Control Room boundary in MODES 1, 2, 3, and 4'' from 24
hours to 14 days for implementation of the Turbine Generator Protection
System Digital Modification currently scheduled during the eleventh
refueling outage for Unit 1 (1RF11) and the ninth refueling outage for
Unit 2 (2RF09). The description of CONDITION E would also be revised
for implementation of this modification.
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. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
1. Do the proposed changes involve a significant increase the
probability or consequences of an accident previously evaluated?
Response: No.
[[Page 29802]]
This is a revision to the Technical Specifications for the CREFS
which is a mitigation system designed to minimize in leakage and to
filter the Control Room atmosphere to protect the operator following
accidents previously analyzed. An important part of the system is
the Control Room boundary. The Control Room boundary integrity is
not an initiator or precursor to any accident previously evaluated.
Therefore, the probability of any accident previously evaluated is
not increased. The analysis of the consequences of analyzed accident
scenarios under the Control Room breach conditions along with the
compensatory actions for restoration of Control Room integrity
demonstrate that the consequences of any accident previously
evaluated are not increased. Therefore, it is concluded that this
change does not significantly increase the probability or
consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change will not impact the accident analysis. The
change will not alter the requirements of the CREFS or its function
during accident conditions. The administrative controls and
compensatory actions will ensure the CREFS will perform its safety
function. No new or different accidents result from the revised
Completion Time or the restated TS Condition E. The change does not
involve a physical alteration of the plant (i.e., no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operation. 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 does not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not affected by these changes. The proposed change will not
result in plant operation in a configuration outside the design
basis for an unacceptable period or time without compensatory
actions and administrative controls. The proposed change does not
affect systems that respond to safely shutdown the plant and to
maintain the plant in a safe shutdown condition. Therefore the
proposed change does not involve a reduction in a margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92'') are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Allen G. Howe.
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].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of application for amendment: September 15, 2004.
Brief description of amendment: The amendment deleted the Technical
Specification (TS) requirements related to hydrogen recombiners and
hydrogen/oxygen monitors. The TS changes are consistent with the
revision of Title 10, Code of Federal Regulations, Section 50.44,
``Standards for Combustible Gas Control System in Light-Water-Cooled
Power Reactors,'' that became effective on October 16, 2003; and
Revision 1 of the NRC-approved Industry/Technical Specifications Task
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
447, ``Elimination of Hydrogen Recombiners and Change to Hydrogen and
Oxygen Monitors.''
Date of issuance: April 28, 2005.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 164.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 01, 2005 (70
FR 5235). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 28, 2005.
No significant hazards consideration comments received: No.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: December 1, 2004.
Brief description of amendments: The amendments eliminate the
requirements to submit monthly operating reports and occupational
radiation exposure reports.
Date of issuance: May 9, 2005.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 272 and 249.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5236).
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated May 9, 2005.
No significant hazards consideration comments received: No.
[[Page 29803]]
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: December 6, 2004.
Brief description of amendment: The amendment deleted the
requirements to submit monthly operating reports and occupational
radiation exposure reports.
Date of issuance: April 28, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 166.
Facility Operating License No. NPF-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5236).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 28, 2005.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: July 19, 2004, as supplemented
by letters dated March 8 and March 22, 2005.
Brief description of amendments: The amendments revised the
Technical Specifications (TS) 3.8.4, ``DC Sources--Operating'' and TS
3.8.6, ``Battery Cell Parameters'' to allow for the replacement of the
existing nickel-cadmium diesel generator batteries with conventional
lead-acid batteries.
Date of issuance: April 27, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance April 27, 2005.
Amendment Nos.: 223 and 218.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 21, 2004 (69
FR 76488). The supplements dated March 8 and March 22, 2005, provided
additional information that clarified the application, did not expand
the scope of the July 19, 2004, application as originally noticed, and
did not change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 27, 2005.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: June 9, 2004, and as
supplemented by letter dated April 1, 2005.
Brief description of amendment: This amendment revises Technical
Specifications (TS) Limiting Condition for Operation (LCO) 3.4.11,
``RCS [Reactor Coolant System] Pressure and Temperature (P/T) Limits,''
to replace the P/T curves for inservice leak and hydrostatic testing,
non-nuclear heating and cooldown, and nuclear heating and cooldown
currently illustrated in TS Figures 3.4.11-1, 3.4.11-2, and 3.4.11-3,
respectively.
Date of issuance: May 12, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 193.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 31, 2004 (69 FR
53102).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 12, 2005.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: December 17, 2004.
Brief description of amendment: The amendment deletes Technical
Specification (TS) 5.6.1, ``Occupational Radiation Exposure Report,''
and TS 5.6.4, ``Monthly Operating Reports.''
Date of issuance: May 3, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment No: 167.
Facility Operating License No. NPF-29: The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 1, 2005 (70 FR
9992).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 3, 2005.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: April 14, 2004, as supplemented
on December 15, 2004.
Brief description of amendment: This amendment eliminates secondary
containment operability requirements when handling sufficiently decayed
irradiated fuel or performing core alterations. The secondary
containment is still required to be operable during operations with the
potential to drain the reactor vessel.
Date of issuance: April 28, 2005.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 215.
Facility Operating License No. DPR-35: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 12, 2004 (69 FR
60679). The December 15, 2004, supplement provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards consideration
determination as published in the Federal Register. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated April 28, 2005.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: July 17, 2004, as supplemented by
letters dated October 18, 2004, February 2, February 21, March 8, and
April 5, 2005.
Brief description of amendment: The amendment revised Technical
Specification (TS) 5.3.1, to allow the use of a limited number of lead
test assemblies, the use of ZIRLOTM as an acceptable fuel
cladding, and to allow a limited substitution of zirconium alloy or
stainless steel filler rods for fuel rods, while relocating the maximum
fuel enrichment from TS 5.3.1 to TS 5.6.1. TS 6.9.1.11.1 is revised to
allow the use of the Westinghouse Nuclear Physics code package and to
incorporate the methodology used to support ZIRLOTM cladding
material. Additionally, the amendment approved the administrative
changes of correcting a referencing report error of the CESEC code and
deleting the TS Index from the TSs.
[[Page 29804]]
Date of issuance: May 9, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 200.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 20, 2004 (69 FR
43460). The supplements dated October 18, 2004, February 2, February
21, March 8, and April 5, 2005, provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination as published in the
Federal Register. The Commission's related evaluation of the amendment
is contained in a Safety Evaluation dated May 9, 2005.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois
Docket Nos. 50-010, 50-237 and 50-249, Dresden Nuclear Power Station,
Units 1, 2 and 3, Grundy County, Illinois
Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 2,
LaSalle County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units
1 and 2, Rock Island County, Illinois
Docket Nos. 50-295 and 50-304, Zion Nuclear Power Station, Units 1 and
2, Lake County, Illinois
Date of application for amendments: October 21, 2004, as
supplemented January 4, 2005.
Description of amendments requests: The amendment deletes the TS
requirements to submit monthly operating reports and annual
occupational radiation exposure reports. The change is consistent with
Revision 1 of NRC-approved Technical Specifications Task Force (TSTF)
Standard Technical Specification Change Traveler, TSTF-369,
``Elimination of Requirements for Monthly Operating Reports and
Occupational Radiation Exposure Reports.'' This TS improvement was
announced in the Federal Register (69 FR 35067) on June 23, 2004, as
part of the Consolidated Line Item Improvement Process (CLIIP).
Date of issuance: April 29, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: Byron Station, Unit 1--142, Unit 2--142; Braidwood
Station, Unit 1--136, Unit 2--136; Dresden Nuclear Power Station, Unit
1--42, Unit 2--214, Unit 3--206; LaSalle County Station, Unit 1--173,
Unit 2--159; Quad Cities Nuclear Power Station, Unit 1--225, Unit 2--
220; Zion Nuclear Power Station, Unit 1--184, Unit 2--171.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72, NPF-77,
DPR-2, DPR-19, DPR-25, NPF-11, NPF-18, DPR-29 and DPR-30: The
amendments revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. Date of initial notice in Federal Register:
April 08, 2005 (70 FR 18061). The notice provided an opportunity to
submit comments on the Commission's proposed NSHC determination. No
comments have been received.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 29, 2005.
Attorney for licensee: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Gene Y. Suh.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendment: October 21, 2004.
Brief description of amendment: The amendments deleted the
Technical Specifications (TSs) 6.9.1.5.a and 6.9.1.6 requirements to
submit monthly operating reports and annual occupational radiation
exposure reports. The change is consistent with Revision 1 of the U.S.
Nuclear Regulatory Commission's Technical Specifications Task Force
(TSTF) Change Traveler, TSTF-369, ``Elimination of Requirements for
Monthly Operating Reports and Occupational Radiation Exposure
Reports.''
Date of issuance: April 29, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 175 and 137.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the TSs.
Date of initial notice in Federal Register: June 23, 2004 (69 FR
35067). This TS improvement was announced in the Federal Register as
part of the Consolidated Line Item Improvement Process. A notice for
these TS changes was announced on April 8, 2005 (70 FR 18059). The
April 8, 2005, notice incorrectly referenced a January 4, 2005,
supplement to the application. This supplement was reference by error.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated April 29, 2005.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: September 10, 2004.
Brief description of amendment: This amendment deletes the
Technical Specifications associated with hydrogen recombiners and
hydrogen monitors.
Date of issuance: April 19, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 135.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 15, 2005 (70
FR 7767). Add the following statement, if appropriate.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 19, 2005.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of application for amendment: January 28, 2004, as
supplemented by letter dated November, 22, 2004.
Brief description of amendment: This amendment revised technical
specifications (TSs) 1.4, ``Frequency,'' 5.5.2, ``Primary Coolant
Sources Outside Containment,'' and 5.5.11, ``Safety Function
Determination Program,'' by adopting three industry-proposed Standard
Technical Specifications (STS) changes, which the Nuclear Regulatory
Commission (NRC) has approved and included in Revision 3 of the STSs.
These changes are Technical Specifications Task Force (TSTF) traveler
numbers 273, 284, and 299. The licensee's request to revise TS 3.3.1.1,
``Reactor Protection System Instrumentation,'' which is associated with
TSTF-264 is addressed by the NRC staff by a separate Safety Evaluation.
Date of issuance: May 12, 2005.
[[Page 29805]]
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 258.
Facility Operating License No. DPR-49: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19571).
The supplemental letter contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 12, 2005.
No significant hazards consideration comments received: No.
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 application for amendments: February 10, 2004.
Brief description of amendments: The amendments (1) extended from 1
hour to 24 hours the completion time (CT) for Condition C of technical
specification (TS) 3.5.1, which defines requirements for the safety
injection accumulators. Condition C of TS 3.5.1 specifies a CT to
restore an accumulator to operable status when it has been declared
inoperable for a reason other than the boron concentration of the water
in the accumulator not being within the required range; (2) deleted
Condition B which permits one or both accumulators to be inoperable, by
removing power to the accumulator isolation valve(s), for maintenance
or testing; (3) modified Condition E to remove reference to Condition
B; and (4) re-lettered the Conditions and Actions to reflect deletion
of Condition B.
Date of issuance: April 28, 2005.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 217, 222.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19573).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 28, 2005.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: May 3, 2004, as supplemented by
letters dated February 4, and March 28, 2005.
Brief description of amendments: The amendments revise the
licensing to define a new hydraulic analysis methodology for
demonstrating functionality of the cooling water (CL) system following
a design-basis seismic event. The seismic analysis methodology for the
CL system is revised to include (1) evaluation of CL system performance
following a seismic event assuming a rupture of a non-seismic pipe at
the worst case location, and (2) application of acceptance criteria
from the American Society of Mechanical Engineers Boiler and Pressure
Vessel Code, Section lll, to demonstrate that the CL system non-seismic
piping will maintain pressure boundary integrity with design-basis
seismic loads.
Date of issuance: May 10, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 169, 159.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Updated Safety Analysis Report.
Date of initial notice in Federal Register: July 6, 2004 (69 FR
40677).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 10, 2005.
No significant hazards consideration comments received: No.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: August 6, 2004, as supplemented
March 14, 2005.
Brief description of amendment: This amendment deletes the
Technical Specification requirements associated with hydrogen
recombiners and hydrogen monitors.
Date of issuance: May 5, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 90.
Facility Operating License No. DPR-18: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 15, 2005 (70
FR 7768). The supplement dated March 14, 2005, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 5, 2005.
No significant hazards consideration comments received: No.
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 application for amendments: December 27, 2004.
Brief description of amendments: The amendments delete TS
5.7.1.1.a, ``Occupational Radiation Exposure Report'' and TS 5.7.1.4,
``Monthly Operating Reports.''
Date of issuance: May 10, 2005.
Effective date: May 10, 2005, to be implemented within 60 days of
issuance.
Amendment Nos.: Unit 2--195; Unit 3--186.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5248). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 10, 2005.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant,
Unit 2, Hamilton County, Tennessee
Date of application for amendment: December 2, 2004, as
supplemented by letters dated February 15, March 9, and April 11, 2005.
Brief description of amendment: The amendment revises portions of
the Sequoyah Unit 2 Technical Specification Surveillance Requirement
4.4.5 to eliminate the requirement to inspect a portion of the tube
within the tubesheet region. This will allow any flaws in the region,
which is no longer inspected, to remain in service.
Date of issuance: May 3, 2005.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 291.
Facility Operating License No. DPR-79: Amendment revises the
technical specifications.
[[Page 29806]]
Date of initial notice in Federal Register: January 18, 2005 (70 FR
2899). The supplemental letters provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 3, 2005.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: October 28, 2004.
Brief description of amendments: This amendment deletes the
Technical Specifications associated with hydrogen recombiners and
hydrogen monitors.
Date of issuance: April 21, 2005.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 117/117.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 15, 2005 (70
FR 7770).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 21, 2005.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room (PDR), located at One White Flint North, Public File Area 01F21,
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly
available records will be accessible from the Agencywide Documents
Access and Management System's (ADAMS) Public Electronic Reading Room
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the PDR Reference
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, the licensee may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene. Requests for a hearing and a petition for leave
to intervene shall be filed in accordance with the Commission's ``Rules
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2.
Interested persons should consult a current copy of 10 CFR 2.309, which
is available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a
request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or a presiding officer designated by the
Commission or by the
[[Page 29807]]
Chief Administrative Judge of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
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\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer or the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: April 18, 2005, as supplemented by
letter dated April 19, 2005.
Description of amendment request: The amendment revises Technical
Specification (TS) 5.5.9, ``Steam Generator (SG) Tube Surveillance
Program,'' to add changes to the SG inspection scope for Wolf Creek
Generating Station for only the current refueling outage 14 and the
subsequent operating cycle. Specifically, the amendment modifies the
inspection requirements for portions of the SG tubes within the hot leg
tubesheet region of the SGs.
Date of issuance: April 28, 2005.
Effective date: Effective the date of issuance, and shall be
implemented before entry into Mode 4 in the restart from the current
Refueling Outage 14.
Amendment No.: 162.
Facility Operating License No. NPF-42: Amendment revises the
technical specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. The Coffey County Republican on April 22 and
26, 2005, and the Emporia Gazette on April 25 and 26, 2005. The notice
provided an opportunity to submit comments on the Commission's proposed
NSHC determination. Comments have been received. The resolution of the
comments, the Commission's related evaluation of the amendment, finding
of exigent circumstances, state consultation, and final NSHC
[[Page 29808]]
determination are contained in a safety evaluation dated April 28,
2005.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Robert A. Gramm.
Dated at Rockville, Maryland, this 16th day of May, 2005.
For the Nuclear Regulatory Commission.
James E. Lyons,
Deputy Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 05-10063 Filed 5-23-05; 8:45 am]
BILLING CODE 7590-01-P