[Federal Register Volume 69, Number 234 (Tuesday, December 7, 2004)]
[Notices]
[Pages 70712-70727]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-26606]
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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
[[Page 70713]]
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 November 5, 2004, through November 24, 2004.
The last biweekly notice was published on November 23, 2004 (69 FRN
68180).
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. (Note:
Public access to ADAMS has been temporarily suspended so that security
reviews of publicly available documents may be performed and
potentially sensitive information removed. Please check the NRC Web
site for updates on the resumption of ADAMS access.) 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/. (Note: Public access to ADAMS has been
temporarily suspended so that security reviews of publicly available
documents may be performed and potentially sensitive information
removed. Please check the NRC Web site for updates on the resumption of
ADAMS access.) If a request for a hearing or petition for leave to
intervene is filed within 60 days, the Commission or a presiding
officer designated by the Commission or by the Chief Administrative
Judge of the Atomic Safety and Licensing Board Panel, will rule on the
request and/or petition; and the Secretary or the Chief Administrative
Judge of the Atomic Safety and Licensing Board will issue a notice of a
hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the
[[Page 70714]]
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 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. (Note: Public access to ADAMS has been
temporarily suspended so that security reviews of publicly available
documents may be performed and potentially sensitive information
removed. Please check the NRC Web site for updates on the resumption of
ADAMS access.) If you do not have access to ADAMS or if there are
problems in accessing the documents located in ADAMS, contact the NRC
PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to
[email protected].
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: October 20, 2004.
Description of amendment request: The proposed amendment would
revise the containment hatch closure requirement in the Technical
Specifications (TSs) during fuel handling and refueling operations.
Specifically, the requirement of TS 3.8.6 that the containment
equipment hatch remain closed with a minimum of 4 bolts securing it in
place is replaced with the requirement that the equipment hatch be
capable of being closed during fuel handling and refueling operations.
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 change is only related to a postulated fuel
handling accident inside the Reactor Building occurring during fuel
loading and refueling activities. The proposed change does not
increase the probability of a fuel handling accident in that the
proposed change deals with the results of such an accident, not the
cause of such an accident. The proposed change does not increase the
consequences of an accident previously evaluated in that the TMI
Unit 1 Alternative Source Term has been previously reviewed and
approved by the NRC [Nuclear Regulatory Commission], and this
proposed change is consistent with the assumptions of [that]
previous analysis. The Alternative Source Term analysis for the Fuel
Handling Accident [i]nside the Reactor Building takes no credit for
the closure of the containment equipment hatch opening or for a
filtered release. Previous analyses of external events were reviewed
and the proposed [change does] not affect the conclusions of these
analyses. Therefore the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not affect nor create a different
[kind] of fuel handling accident. The proposed change is consistent
with the existing licensing basis accident analysis for a postulated
fuel handling accident inside containment during fuel loading and
refueling operations. The proposed change does not involve any
structure, system, or component relied upon to mitigate any design
basis accident. The revised operations are consistent with the fuel
handling accident analysis. 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 amendment involve a significant reduction
in a margin of safety?
Response: No.
Previously approved analysis demonstrates that the resultant
dose consequences are well within the appropriate acceptance
criteria. The proposed change is bounded by the previously approved
analysis, and thus the margin of safety, as defined by 10 CFR 50.67
and Regulatory Guide 1.183, is maintained. Maintaining the
capability to close the containment equipment hatch opening
following an evacuation of the containment would further reduce the
dose consequences in the event of a fuel handling accident inside
containment and provides additional margin to the calculated doses.
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
[[Page 70715]]
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: Thomas S. O'Neill, Associate General
Counsel, AmerGen Energy Company, LLC, 4300 Winfield Road, Warrenville,
IL 60555.
NRC Section Chief: Richard J. Laufer.
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 amendment request: July 13, 2004.
Description of amendment request: The proposed amendment would
revise the fire protection license condition consistent with the
guidance provided in Generic Letter 88-12, ``Removal of Fire Protection
Requirements from 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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed license amendment is an administrative change. The
proposed change will revise the fire protection license condition
consistent with the guidance provided in Generic Letter 88-12. This
revision to the fire protection license condition was to be made at
the time the fire protection requirements were relocated from the
Technical Specifications to licensee controlled documents. However,
this change was not requested, nor granted in License Amendment
Request dated December 4, 1996, approved in Amendment Nos. 227 and
201. Therefore, the necessary change was not reflected in the
Operating Licenses.
This administrative request does not impact the probability or
consequences of an accident previously evaluated. The incorporation
of the requested change requires that an evaluation be performed to
determine the need for prior NRC approval for changes to the Fire
Protection Program. Changes to administrative programs will result
from the addition of this condition in the Operating License.
However, no changes to the facility or the way it is operated are
expected to result from this change.
Therefore, this proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different kind
of accident from any accident previously evaluated.
The proposed change is administrative. This request does not
involve a change in the operation of the plant, and no new accident
initiation mechanism is created by the proposed change, nor does the
change involve a physical alteration of the plant.
Therefore, the proposed change does not create the possibility
of a new or different [kind] of accident from any accident
previously evaluated.
3. Would not involve a significant reduction in [a] margin of
safety.
The fire protection requirements were removed from the Technical
Specifications in accordance with Generic Letter 88-12 in Amendment
Nos. 227 and 201, with the exception of the change to the Operating
License's fire protection license condition. The proposed
administrative change will require an evaluation be performed to
determine the need for prior NRC approval for changes to the Fire
Protection Program. No margin of safety is impacted by the proposed
administrative change.
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: James M. Petro, Jr., Esquire, Counsel,
Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor,
Baltimore, MD 21202.
NRC Section Chief: Richard J. Laufer.
Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423,
Millstone Power Station, Unit No. 3, New London County, Connecticut.
Date of amendment request: September 7, 2004.
Description of amendment request: The proposed changes would allow
performance of testing for nozzle containment spray blockage to be
based on the occurrence of activities that could cause nozzle blockage
rather than a fixed periodic basis. Currently, the testing for nozzle
blockage is performed every 10 years and Dominion Nuclear Connecticut,
Inc. (DNC) proposes to change this frequency to ``following maintenance
that could cause nozzle blockage''. In addition, specific details
limiting the testing method to an air or smoke test that are currently
part of the surveillance requirements would be removed. The Technical
Specification Bases section would be updated with applicable spray
nozzle testing information and will be expanded to include visual
inspection.
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 spray nozzles and the associated containment spray systems
are designed to perform accident mitigation functions only. The QSS
[quench spray system] and RSS [recirculation spray system] and
associated components are not considered as initiators of any
analyzed accidents. The proposed change does not modify any plant
equipment and only changes the frequency for performance of a
surveillance test which does not impact any failure modes that could
lead to an accident. Removing the testing details from the
surveillance does not change the ability of the spray nozzles to
function as assumed and therefore there is no affect on the
consequence of any accident. Also the proposed change does not
impact the capability of the QSS and RSS to perform accident
mitigation functions and therefore does not impact the consequences
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 QSS and RSS are not being physically modified and there is
no impact on the capability of the systems to perform accident
mitigation functions. No system setpoints are being modified and no
changes are being made to the method in which borated water is
delivered to the spray nozzles. The testing requirements imposed by
this proposed change to check for nozzle blockage following
activities that could cause nozzle blockage do not introduce new
failure modes for the system. By removing the testing details from
the surveillance requirement, additional flexibility in the testing
methodology is allowed for verifying the nozzles are unobstructed
and assists in ensuring operability of the systems. 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 proposed change does not change or introduce any new
setpoints at which mitigating functions are initiated. No changes to
the design parameters of the QSS and RSS are being proposed. No
changes in system operation are being proposed by this change that
would impact an established safety margin. The proposed change
modifies the frequency for verification of nozzle operability in
such a way that continued high confidence exists for the containment
spray systems to functions as designed. In addition, removing
specific testing details from the surveillance does not affect the
ability of the
[[Page 70716]]
spray nozzles to function as designed. Therefore, based on the
above, the proposed amendment does not involve a significant
reduction in a margin of safety.
In summary, DNC concludes that the proposed amendment does not
represent a significant hazards consideration under the standards
set forth in 10 CFR 50.92(c).
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: Daniel S. Collins, Acting Section Chief.
Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear
Generating Unit No. 2, Westchester County, New York
Date of amendment request: November 1, 2004.
Description of amendment request: The proposed amendment would
allow the use of a new gantry crane as part of the cask handling system
in the fuel storage building (FSB) for moving spent fuel casks up to
110 tons into and out of the spent fuel pit.
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 amendment would allow the use of the new [IP2]
FSB gantry crane for loads up to 110 tons, and the new crane will
prevent the load from being dropped given a single malfunction or
failure of a portion of the crane. The handling of a loaded spent
fuel cask is below the maximum load that the crane is designed to
handle.
This change does not increase the probability of an accident
previously evaluated because the probability of a load drop is
eliminated. The new crane system is designed in accordance with
NUREG-0554 and Ederer's Generic Licensing Topical Report EDR-1 (NP)-
A, that if a portion of the crane lifting devices malfunctions or
fails, the load will move a limited distance downward prior to
backup restraints becoming engaged. The change does not increase the
consequences of an accident.
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. The process for transporting a cask with the new crane is
limited from the FSB truck bay floor to the cask pit area of the
spent fuel pool. Once a cask is loaded with spent fuel, it is lifted
from the spent fuel pit, and lowered into the truck bay. The cask is
never carried over spent fuel in the spent fuel pit.
Therefore, the change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. The [IP2] FSB gantry crane has been installed to comply with
the single-failure-proof requirements of NUREG-0554 and NRC-approved
Ederer Topical Report EDR-1, Revision 3, dated October 8, 1982. The
installation provides additional load carrying capability up to 110
tons and additional safety features to prevent a cask drop. The
safety margins provided by the new crane prevent failure of the
crane or any lifting devices associated with it. The implementation
of NUREG-0612 general guidelines for the FSB gantry crane provides
further assurance that safe load paths, procedures, crane operator
training, and crane inspection and maintenance activities will be
established to ensure crane operation is performed in a consistently
reliable manner.
Therefore, operation of the facility in accordance with 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: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 29, 2004.
Description of amendment request: The proposed change to the
Technical Specification (TS) assures that sufficient fuel oil
inventories are available in the Emergency Diesel Generator (DG) Fuel
Oil Storage Tank (FOST) to support the Extended Power Uprate (EPU)
consistent with the current licensing basis.
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, which accounts for the fuel oil consumption
related to the EPU, will revise the minimum TS volumes associated
with the DG FOST. The change continues to assure that each DG can
provide on-site power in the event of an accident and thereby assist
in the mitigation of the accident.
The proposed change to the five day full load fuel oil volume
results in a usable volume 37,000 gallons of fuel oil. The proposed
change removes the unusable volume (760 gallons) and other
conservatism (240 gallons) that were included in the current TS. The
fuel oil volume continues to allow for a runtime of 5 days at full
load with the removal of this conservatism.
These changes will not affect the capability of the AC
[alternate current] Sources to power the systems required to safely
shutdown the plant. The proposed changes are not accident initiators
nor do they adversely affect accident initiators or precursors.
These changes do not affect the mitigation of any accident nor do
they adversely affect structures, systems, or components that are
utilized for the mitigation of any analyzed events. The proposed
changes will have no affect on the radiological consequences of any
accident.
The proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in the
evaluation of radiological consequences.
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.
Fuel oil is not an accident initiator. Therefore, the
possibility of a new or different kind of accident will not be
created in relationship to the proposed changes to the TS. No
modifications are proposed to the existing fuel oil storage system
that would alter the design function or the ability of the DG to
perform its safety function.
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 to the 7-day time dependent fuel oil volume
results in an increase in volume to accommodate fuel oil consumption
needed to support the EPU.
The reduced volume associated with the 5-day full load volume is
equivalent to less than one hour of runtime and does not result in a
significant reduction in a margin of safety because the
calculational method results in a conservative estimate of the
amount of fuel that would be needed during a design bases accident.
The proposed change does not result in a change of the design
bases for the DG or its support systems. The system will continue to
provide a reliable source of power for safe shutdown of the reactor,
assuming the single failure of one of the DGs. Independence,
redundancy, and testability are maintained such that the required
safety function can be performed by either DG train.
[[Page 70717]]
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: N.S. Reynolds, Esquire, Winston & Strawn,
1400 L Street NW., Washington, DC 20005-3502.
NRC Section Chief: Michael K. Webb, Acting.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3 (Waterford 3), St. Charles Parish, Louisiana
Date of amendment request: November 5, 2004.
Description of amendment request: Waterford 3 Technical
Specification (TS) currently requires all the Dry Cooling Tower (DCT)
fans with cooling coils under the missile grating to be operable during
a tornado watch. If one (or more) of these DCT fans is inoperable
during a tornado watch, it is required to be restored to operable
status within one hour or place the plant in Hot Standby within 6
hours. The purpose of this TS change is to allow the plant to take
credit for the DCT fans that are not under the missile grating to meet
the fan requirements specified in TS Table 3.7-3. In addition, the
proposed change will delete the requirement to monitor ambient
temperature conditions when the DCT fan is inoperable on an inoperable
train of the Ultimate Heat Sink (UHS).
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 will delete the requirement to have all the
DCT fans with cooling coils under the missile grating operable
during a tornado watch. It has been determined (using tornado
missile strike probability methodology--TORMIS) that the probability
of damage to the DCT components not under the missile grating (fans,
motors, associated conduits, electrical boxes, and cooling coils) is
acceptably low. With respect to the probability of occurrence or the
consequences of an accident previously analyzed in the FSAR [Final
Safety Analysis Report], the possibility of a tornado reaching
Waterford 3 and causing damage to plant systems, structures and
components, including the DCT fans, is a design basis event
considered in the FSAR. The probability of a tornado-generated
missile strike on the DCT components was analyzed using the NRC
[Nuclear Regulatory Commission] Staff approved probability method
TORMIS. TORMIS showed that the change from essentially relying on
DCT fans with cooling coils under the missile grating to relying on
all operable DCT fans during a tornado watch is acceptable and
represents an acceptably low probability of occurrence of tornado
generated missile strikes on the DCTs. On this basis, the proposed
change is not considered to constitute a significant increase in the
probability of occurrence or the consequences of an accident.
The proposed change to TS Action 3.7.4.d eliminates an
unnecessary requirement, to determine ambient conditions and verify
compliance with TS Table 3.7-3, when an Ultimate Heat Sink (UHS) fan
is inoperable due to its associated train of UHS being inoperable.
The determination of ambient temperature conditions and validation
of the required number of fans based on the temperature will
continue to be required when an UHS fan is inoperable and the
associated train of UHS is operable. The UHS fans will not dissipate
the required heat load when the associated train of UHS is
inoperable, assuming the coincident ambient wet bulb temperature (78
[deg]F) at the historically highest ambient dry bulb temperature
(102 [deg]F). This change represents a burden reduction and has no
impact on plant safety. This change also does not impact the
initiators or mitigation of any design basis event.
The proposed revision to TS Table 3.7-3 ensures consistency with
the revisions to the TS Actions. This change is administrative and
has no impact on the initiators or the mitigation of accidents
previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change will delete the requirement to have all the
DCT fans with cooling coils under the missile grating operable
during a tornado watch. It has been determined that the probability
of damage to the DCT components not under the missile grating is
acceptably low. A tornado at Waterford 3 is a design basis event
considered in the FSAR. Therefore, the change will not contribute to
the possibility of or be the initiator for any new or different kind
of accident, or occur coincident with any of the design basis
accidents in the FSAR. The low probability threshold established for
tornado missile damage to system components is consistent with these
assumptions.
The proposed change to TS Action 3.7.4.d eliminates an
unnecessary requirement, to determine ambient conditions and verify
compliance with TS Table 3.7-3, when an Ultimate Heat Sink (UHS) fan
is inoperable due to its associated train of UHS being inoperable.
The determination of ambient temperature conditions will continue to
be required when an UHS fan is inoperable with the associated train
of UHS operable. There are no plant modifications or design changes
proposed.
The proposed revision to TS Table 3.7-3 ensures consistency with
the revisions to the TS Actions. This is an administrative change.
The above changes also do not have any impact on plant systems
nor do they have any impact on the way plant systems are operated.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident.
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. The existing licensing basis for Waterford 3 with
respect to the design basis event of a tornado reaching the plant,
generating missiles, and directing them toward the DCT components is
to provide positive missile barriers. The basis for the proposed
change recognizes that there is a low probability, below an
established acceptance limit, that a tornado missile will strike DCT
components. The change from essentially relying on DCT fans with
cooling coils under the missile grating to relying on all operable
DCT fans during a tornado watch is acceptable and represents an
acceptably low probability of occurrence of tornado generated
missile strikes on the DCTs. Therefore, this change is not
considered to constitute a significant decrease in the margin of
safety.
The proposed change to TS Action 3.7.4.d eliminates an
unnecessary requirement, to determine ambient conditions and verify
compliance with TS Table 3.7-3, when an Ultimate Heat Sink (UHS) fan
is inoperable due to its associated train of UHS being inoperable.
The determination of ambient temperature conditions will continue to
be required when an UHS fan is inoperable with the associated train
of UHS operable. When the UHS is not available, the fans cannot
dissipate the required heat load, assuming the coincident ambient
wet bulb temperature (78 [deg]F) at the historically highest ambient
dry bulb temperature (102 [deg]F). Therefore, it is not necessary to
monitor ambient temperature and ensure the fan requirements of TS
Table 3.7-3 are met when the UHS train is inoperable. This change
represents an operational burden reduction and has no impact on
plant safety.
The proposed revision to TS Table 3.7-3 ensures consistency with
the revisions to the TS Actions. These changes are administrative
and have no impact on the operation of the plant, mitigation of
analyzed events, or plant safety,
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, Entergy concludes that the proposed
amendment presents no significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
[[Page 70718]]
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: N.S. Reynolds, Esquire, Winston & Strawn
1400 L Street NW., Washington, DC 20005-3502.
NRC Section Chief: Michael K. Webb, Acting.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois; Docket
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and
2, Rock Island County, Illinois
Date of application for amendment request: February 27, 2004, as
supplemented by letter dated October 11, 2004.
Description of amendment request: The proposed amendments would
revise the Dresden Nuclear Power Station and Quad Cities Nuclear Power
Station technical specifications (TS) to add the Oscillation Power
Range Monitor (OPRM) instrumentation to 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 change involve a significant increase in
the probability or consequences of an accident previously evaluated?
This proposed change has no impact on any of the existing
neutron monitoring functions. It activates the OPRM scram function
and updates the TS to add the OPRM-related functions.
Activation of the OPRM scram function will replace the current
methods that require operators to insert an immediate manual reactor
scram in the reactor operating region where thermal hydraulic
instabilities could potentially occur. While this region will
continue to be avoided during normal operation, certain transients,
such as a reduction in reactor recirculation flow, could place the
reactor in this region. Operation in this region, with the OPRM
instrumentation scram function activated would no longer require an
immediate manual scram and thus may potentially cause a marginal
increase in the probability of occurrence of an instability event.
This potential increase in probability is acceptable because the
OPRM function will automatically detect the instability condition
and initiate a reactor scram before the Minimum Critical Power Ratio
(MCPR) Safety Limit is reached. Consequences of the potential
instability event are reduced because of the more reliable automatic
detection and suppression of an instability event, and the
elimination of dependence on the manual operator actions. Operators
will continue to monitor for indications of thermal hydraulic
instability when the reactor is operating in regions of potential
instability as a backup to the OPRM instrumentation.
The potential for spurious reactor scrams has been evaluated.
Operating experience with the OPRM has not resulted in the
generation of any spurious reactor scram signals.
Therefore, the proposed changes do not involve a significant
increase in the probability of 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 changes replace procedural actions that were
established to avoid operating conditions where reactor
instabilities might occur with an NRC approved automatic detect and
suppress function (i.e., OPRM).
Potential failures in the OPRM trip function could result in
either failure to take the required mitigating action or an
unintended reactor scram. These are the same potential effects of
failure of the operator to take the appropriate action under the
current procedural actions. The effects of failure of the OPRM
equipment are limited to reduced or failed mitigation, but such
failure cannot cause an instability event or other type of accident.
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?
The OPRM trip function is being implemented to automate the
detection and subsequent suppression of an instability event prior
to exceeding the MCPR Safety Limit. The OPRM trip provides a trip
output of the same type as currently used for the APRM [Average
Power Range Monitor]. Its failure modes and types are identical to
those for the present APRM output. Since the MCPR Safety Limit will
not be exceeded as a result of an instability event following
implementation of the OPRM trip function, it is concluded that the
proposed change does not reduce the 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: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Gene Y. Suh.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of amendment request: October 4, 2004.
Description of amendment request: The proposed change requests
approval to apply the Westinghouse best-estimate loss-of-coolant
accident (BELOCA) analysis methodology to Beaver Valley Power Station,
Unit Nos. 1 and 2, and requests an amendment of the related Technical
Specifications. The BELOCA methodology has previously been approved on
a generic basis by the NRC as presented in Topical Report WCAP-12945-P-
A, Volume 1 (Revision 2) and Volumes 2 through 5 (Revision 1), ``Code
Qualification Document for Best-Estimate LOCA Analysis,'' March 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?
No. No physical changes are required as a result of implementing
best-estimate large break loss of coolant accident (LOCA)
methodology and associated Technical Specification changes. The
plant conditions used in the analysis are bounded by the design
conditions for all equipment in the plant. Therefore, there will be
no increase in the probability of a LOCA. The consequences of a LOCA
are not being increased, since it is shown that the emergency core
cooling system is designed so that its calculated cooling
performance conforms to the criteria contained in 10 CFR 50.46,
Paragraph b. No other accident is potentially affected by this
change.
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 analyzed?
No. There are no physical changes being made to the Beaver
Valley Power Station units. No new modes of plant operation are
being introduced. The parameters used in the analysis are within the
design limits of the existing plant equipment. All plant systems
will perform as designed during the response to a potential
accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously analyzed.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
[[Page 70719]]
No. It has been shown that the methodology used in the analysis
would more realistically describe the expected behavior of plant
systems during a postulated LOCA. Uncertainties have been accounted
for as required by 10 CFR 50.46. A sufficient number of LOCAs with
different break sizes, different locations and other variations in
properties are analyzed to provide assurance that the most severe
postulated LOCAs are addressed. It has been shown by analysis that
there is a high probability that all criteria contained in 10 CFR
50.46, Paragraph b are met.
Therefore the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: October 15, 2004.
Description of amendment request: The licensee proposed to revise
Section 1.7, regarding the definition of ``Instrument Channel
Calibration,'' of the Technical Specifications by incorporating the
additional guidance for instrument channels containing resistance
temperature detector (RTD) and thermocouple sensors provided by the
``Standard Technical Specifications, General Electric Plants, BWR
[Boiling-Water Reactor]/4 Specifications,'' NUREG-1433, Revision 3. The
revised definition would permit in place qualitative assessment of the
RTDs and thermocouples, and to allow a signal to be injected downstream
of the sensor for the purpose of calibrating the remainder of the
channel.
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 revises the definition of Instrument Channel
Calibration to allow RTD and thermocouple sensors to be
qualitatively assessed with the remainder of the channel being
calibrated normally. Instrument channel calibration is not an
initiator of any accident previously evaluated. Furthermore, the
proposed change will not affect the ability of the channel being
calibrated to respond as assumed in any accident previously
evaluated. The qualitative evaluation of sensor behavior for non-
adjustable sensors will provide an accurate indication of sensor
operation and will assure that portion of the channel is operating
properly, ensuring that the consequences of an accident will remain
as previously evaluated. Therefore, the proposed Technical
Specification 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 change revises the definition of Instrument Channel
Calibration to allow RTD and thermocouple sensors to be
qualitatively assessed with the remainder of the channel being
calibrated as at present. 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. The proposed change also does not adversely
affect the operation or operability of existing plant equipment.
Therefore, operation of the facility in accordance with the proposed
amendment would 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 involves the definition of Instrument
Channel Calibration to allow RTD and thermocouple sensors to be
qualitatively assessed with the remainder of the channel being
calibrated normally. The proposed change to the Instrument Channel
Calibration definition does not alter the ability of a channel to
respond as designed or as assumed in the safety analyses. 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: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Richard J. Laufer.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: October 22, 2004.
Description of amendment request: The licensee proposed to revise
the Technical Specifications (TSs), Section 5.0, ``Design Features,''
by relocating the information to the Updated Final Safety Analysis
Report (UFSAR). Specifically, the amendment would relocate these
Sections: 5.3, ``Reactor Vessel,'' 5.4, ``Containment,'' and 5.6,
``Seismic Design.'' The licensee stated that such information does not
meet the criteria of 10 CFR 50.36(c)(4) for inclusion in the TSs. The
information to be relocated to the UFSAR already exists in the UFSAR,
and will continue to be controlled by 10 CFR 50.59 and 10 CFR 50.71(e).
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 relocates certain design details from the TS
to the UFSAR, where the information already exists. The UFSAR is
maintained in accordance with 10 CFR 50.71(e). Any future change to
these design details as described in the UFSAR will be evaluated per
the requirements of 10 CFR 50.59 to assure that the change does not
result in more than a minimal increase in the probability or
consequences of an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
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. The
proposed change will not impose or eliminate any requirements, and
adequate control of the information will be maintained in accordance
with applicable regulatory requirements.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change has no impact on any analysis assumptions.
The design details that
[[Page 70720]]
are being removed from the TS already exist in the UFSAR. Any future
change to these design details described in the UFSAR will be
evaluated per the requirements of 10 CFR 50.59 to assure that the
change does not result in a design basis limit [or] a fission
product barrier being exceeded or altered.
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: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Richard J. Laufer.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of amendment request: October 14, 2004.
Description of amendment request: The proposed changes correct
administrative errors in Technical Specifications 3.10.i and 6.9.a.4.A.
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?
NMC [Nuclear Management Company, the licensee] Response for
Proposed Change to TS 3.10.i: No. The NMC has reviewed the proposed
change in accordance with the provisions of 10 CFR 50.92 to show no
significant hazards exist. This change is being proposed to correct
an administrative error that currently exists within the KNPP
[Kewaunee Nuclear Power Plant] Technical Specifications; therefore
it would not have an affect on the probability of an accident
previously evaluated.
NMC Response for Proposed Change to TS 6.9.a.4.A: No. The NMC
has reviewed the proposed change in accordance with the provisions
of 10 CFR 50.92 to show no significant hazards exist. This change is
being proposed to correct an administrative error that currently
exists within the KNPP Technical Specifications; therefore it would
not have an affect on the probability 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?
NMC Response for Proposed Change to TS 3.10.i: No. The proposed
change does not alter plant configuration, operating setpoints, or
overall plant performance. Therefore, the proposed change would not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
NMC Response for Proposed Change to TS 6.9.a.4.A: No. The
proposed change does not alter plant configuration, operating
setpoints, or overall plant performance. Therefore, the proposed
change would 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?
NMC Response for Proposed Change to TS 3.10.i: No. The proposed
change does not involve a significant reduction in a margin of
safety. Inclusion of the omitted word ``and'' in TS 3.10.i will
enhance the margin of safety.
NMC Response for Proposed Change to 6.9.a.4.A: No. The proposed
change does not involve a significant reduction in a margin of
safety. Correction of the references in TS Section 6.9.a.4.A will
enhance 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: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Section Chief: L. Raghavan.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California
Date of amendment request: July 9, 2004.
Description of amendment request: The Humboldt Bay Power Plant
(HBPP), Unit 3, is a decommissioning nuclear power plant that was
permanently shutdown in July 1976. In December of 2003, Pacific Gas and
Electric (PG&E or the licensee) applied for a license to store its
spent fuel in an onsite dry cask independent spent fuel storage
installation (ISFSI). Moving the spent fuel to an ISFSI would permit
the licensee to begin significant decommissioning activities. The
licensee has chosen to use a Holtec HI-STAR HB spent fuel cask handling
system involving a spent fuel multipurpose canister and overpack. To
facilitate spent fuel transfer from the HBPP spent fuel pool to the
ISFSI, the licensee will also need to install a new crane that can be
used to lift the cask handling system loaded with spent fuel
assemblies. The licensee states it will be able to satisfy the
applicable guidance of NUREG-0612, ``Control of Heavy Loads at Nuclear
Power Plants,'' and NUREG-0554. ``Single-Failure Proof Cranes for
Nuclear Power Plants,'' in performing the necessary movement of the
HBPP spent fuel to dry cask storage. The licensee has requested a
license amendment that approves the use of the crane and associated
changes to the HBPP Defueled Safety Analysis Report (DSAR) along with
analyses, design, and procedural changes required to implement transfer
of the spent fuel from the spent fuel pool to the ISFSI.
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?
No. With the HI-STAR HB System and the associated design and
handling procedures, all cask drops and other events, which could
damage other spent fuel, have been precluded through the robust
handling systems, and mechanical arrangement that preclude crane
movement over spent fuel, meeting the guidelines of NUREG-0612.
Revisions of the HBPP procedures implementing the control of heavy
loads ensures that PG&E will meet the NUREG-0612 guidelines and will
protect the fuel storage locations and the new HI-STAR HB System
loading/unloading activities. As a result of this design approach, a
cask-handling accident that results in a significant offsite
radiological release is not considered credible as demonstrated by
the probabilistic evaluation that was performed using the guidelines
of NUREG-0612 Appendix B and updated information from NUREG-1774
[``A Survey of Crane Operating Experience at U.S. Nuclear Power
Plants from 1968 through 2002.'']
Other HBPP licensing-basis events, such as the drop of a spent
fuel assembly, have not been affected by these changes and remain
bounding events for potential radiological consequences.
The proposed design of the dry cask system, the handling system,
and associated procedural controls provide assurance that: (1)
Operational errors and mishandling events, and (2) support system
malfunctions will not result in an increase in the probability or
consequence of an accident previously analyzed.
The proposed changes to use the Holtec HI-STAR HB system have
been evaluated for seismic events and tornado missile impacts and it
has been determined that these changes will not result in an
increase in the probability or consequences of an accident
previously evaluated. The Fire Protection Program will ensure that
the combustible materials are properly controlled such that the
total combustibles meet the current program commitments. Therefore,
the
[[Page 70721]]
proposed changes do not involve a significant increase in the
probability or consequences of an accident.
2. Does the proposed amendment create the possibility of a new
or different type of accident from any accident previously
evaluated?
No. The engineering design measures and the handling procedures
preclude the possibility of new or different kinds of accidents.
Damage to 10 CFR 50 structures, systems, and components from the
cask handling and associated activities, and events resulting from
possible damage to contained fuel have been considered. Both the
types of accidents and the results remain within the envelope of
existing HBPP DSAR licensing basis analyses, as demonstrated by the
PG&E and Holtec analyses.
The rupture of multipurpose canister (MPC) dewatering, forced
helium dehydration or related closure system lines or the
malfunction of equipment during cask handling operations resulting
in radiological consequences are bounded by the HBPP DSAR fuel-
handling accident analysis.
Other design considerations, such as spent fuel pool (SFP)
thermal, water chemistry and clarity, criticality, and structural,
were evaluated and determined not to introduce the possibility of a
new or different kind of accident from any previously evaluated.
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 amendment involve a significant reduction
in a margin of safety?
No. With the Holtec HI-STAR HB System, and the associated design
and handling procedures, cask drops and other events have been
precluded through robust load handling systems, providing defense-
in-depth as described in NUREG-0612. Cask tipovers, while not
considered credible, are shown to be below the 60g limit, preventing
damage to the contained fuel assemblies (and associated structures),
and meeting the analysis guidelines of NUREG-0612. As the existing
licensing basis assumes a nonmechanistic drop damaging the SFP and
all fuel, the result of this design approach with the minimization
of drops and the associated structural challenges assure the margin
of safety has been maintained.
Other HBPP licensing-basis events, such as the drop of a spent
fuel assembly, have not been affected by these changes and remain
bounding events. Revision of HBPP procedures implementing the
control of heavy loads to incorporate the additional restrictions on
heavy loads movement will not affect the procedures or methodology
used and will, therefore, not affect margins.
Adverse effects from seismic events and/or cask drops or
tipovers have been evaluated, assuring that the fuel, MPC, and
overpack remain within their design bases. Since design basis
criteria are fully satisfied, there is no impact on the margin of
safety.
The Fire Protection Program will continue to ensure that the
combustible materials are properly controlled such that the total
combustibles meet the current program commitments. Thus, there are
no significant reductions in margin of safety associated with these
changes.
Other design considerations, such as SFP thermal, water
chemistry, criticality, and structural, were evaluated and
determined to not involve a reduction in a margin of safety.
Based on the above evaluations, the licensee concludes that the
activities associated with the above changes present no significant
hazards consideration under the standards set forth in 10 CFR 50.92 and
accordingly, a finding by the NRC of no significant hazards
consideration is justified.
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: Richard F. Locke, Esquire, Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Claudia Craig.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: September 8, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification 3.1.8, ``Scram Discharge Volume (SDV)
Vent and Drain Valves,'' to allow a vent or drain line with one
inoperable valve to be isolated instead of requiring the valve to be
restored to Operable status within 7 days.
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 in
license amendment applications in the Federal Register on April 15,
2003 (68 FR 18294). The licensee affirmed the applicability of the
model NSHC determination in its application dated September 8, 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 one valve inoperable instead of requiring the valve to be
restored to operable status within 7 days. 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.
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, Luzerne County, Pennsylvania
Date of amendment request: September 22, 2004.
Description of amendment request: The proposed amendment would
extend
[[Page 70722]]
the validity of the reactor pressure vessel (RPV) pressure-temperature
(P-T) limit curves from May 1, 2005, to May 1, 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. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The evaluation for the Unit 2 P-T limit curves for 32 EFPYs
[effective full-power years] was performed using the approved
methodologies of 10 CFR [Part] 50 Appendix G and Code Case-640. The
curves generated from these methods were approved as Amendment 174
(Ref. 1) and are currently in the Unit 2 TS. These curves ensure the
P-T limits will not be exceeded during any phase of reactor
operation. Resolution of the current industry issues related to
fluence calculation methodology required PPL to limit applicability
of the curves to May 1, 2005 for Unit 2. The proposed change does
not alter any of the technical information shown on the present P-T
curves. The change extends the expiration date for one year while
maintaining the total accumulated exposure well below the 32 EFPY
maximum exposure lifetime limit. Therefore, there is no increase in
the probability or consequences of any previously evaluated accident
as a result of this change.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change involves changing the expiration date on the
Unit 2 P-T limit curves. The change does not affect the present
operating margin in the P-T limit curves for inservice leakage and
hydrostatic pressure testing, non-nuclear heatup and cooldown, and
criticality. Operation in accordance with the present P-T curves,
developed in accordance with the provisions of ASME Code [American
Society of Mechanical Engineers Boiler and Pressure Vessel Code],
Section XI, Appendix G; 10 CFR [Part] 50 Appendix G, and ASME Code
Case-640 provides adequate protection against a non-ductile-type
fracture of the RPV. This proposed change does not create the
possibility of any new or different [kind] of accident. The change
extends the expiration date of the present P-T curves and does not
result in any new or unanalyzed operation of any system or piece of
equipment important to safety.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The technical information contained in the present P-T curves
approved by Amendment 174 (Ref. 1) is not affected by this change.
Extending the expiration date of the curves from May 1, 2005 to May
1, 2006 will not reduce the margin of safety to RPV brittle
fracture.
Since the Unit 2 P-T curves have a maximum lifetime exposure of
32 EFPYs and the anticipated exposure by May 1, 2006 will be well
below the maximum value, the margin of safety is not reduced as the
result of this change in expiration date. Resolution of the current
industry issues related to fluence calculation methodology requires
PPL to limit applicability of the Unit 2 P-T curves to May 1, 2006.
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.
Tennessee Valley Authority (TVA), Docket No. 50-390, Watts Bar Nuclear
Plant, Unit 1, Rhea County, Tennessee
Date of amendment request: September 23, 2004.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to require automatic starting
of the auxiliary feedwater (AFW) pumps upon trip of the Turbine Driven
Main Feedwater (TDMFW) pumps only when one or more of TDMFW pumps are
operating.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR), Section 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?
No. The design basis events which impose AFW safety function
requirements are loss of normal main feedwater, main feedline or
main steamline break, loss of offsite power, loss of coolant
accident, and small break loss of coolant accident. These accident
evaluations assume actuation of AFW occurring due to low-low steam
generator level or a safety injection signal. These signals are
required safety related features unlike start-up of the AFW pumps
due to the trip of both TDMFW pumps which is an anticipatory
function and not required for either transient or accident analyses.
Requiring this function only when the TDMFW pumps are running will
not impact any previously evaluated design basis events. 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?
No. This TS change involves the automatic start of the AFW pumps
when the TDMFW Pumps trip. This change involves a function that is
not a safety related feature and, therefore, is not credited in
either transient or accident analyses. Since this change only
affects the point at which this trip function needs to be operable
and does not affect the function that actuates AFW due to low-low
steam generator level or a safety injection signal, it will not be
an initiator to a new or different kind of accident from any
accident previously evaluated.
3. Does the proposed change involve a significant reduction in
margin of safety?
No. This TS change involves the automatic start of the AFW pumps
when the TDMFW pumps trip which is not a safety related plant
function. This change does not change any values or limits involved
in a safety related 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Michael L. Marshall, Jr.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: October 27, 2004.
Description of amendment request: The amendment would revise
Technical Specification (TS) 3.7.3, ``Main Feedwater Isolation Valves
(MFIVs),'' to add the main feedwater regulating valves (MFRVs) and the
associated MFRV bypass valves (MFRVBVs). In addition, the allowed
outage time, or completion time, for inoperable MFIVs would be
extended.
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 add the MFRVs and MFRVBVs to TS 3.7.3 and
extend the
[[Page 70723]]
Completion Time for one or more MFIVs inoperable from 4 hours to 72
hours. Extending the Completion Time is not an accident initiator
and thus does not change the probability that an accident will
occur. However, it could potentially affect the consequences of an
accident if an accident occurred during the extended unavailability
of the inoperable MFIV. The increase in time that the MFIV is
unavailable is small and the probability of an event occurring
during this time period which would require isolation of the MFW
[main feedwater] flow paths is low. Moreover, the redundancy
provided by the MFRVs and MFRVBVs, which have the same actuation
signals and closure time requirements as the MFIVs, provides
adequate assurance that automatic feedwater isolation will occur if
called upon.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Closure of the MFIVs is required to mitigate the consequences of
the Main Steam Line Break and Main Feedwater Line Break accidents.
The MFRVs and MFRVBVs provide a diverse backup to this function.
[The extended Completion Time for inoperable MFIVs is not an
accident initiator.] The proposed changes do 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 changes do not revise any Technical Specification
[Safety] Limit or accident analysis assumption. Therefore, [they do]
not involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Robert A. Gramm.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: October 27, 2004.
Description of amendment request: The amendment would delete or
revise license conditions in the operating license for the Callaway
Plant because the requirements are either obsolete or adequately
described elsewhere. The amendment would also revise Technical
Specification Tables 5.5.9-2, ``Steam Generator Tube Inspection,'' and
5.5.9-3, ``Steam Generator Repaired Tube Inspection,'' to delete the
requirement to notify the NRC pursuant to 10 CFR 50.72(b)(2) if the
steam generator tube inspection results in a C-3 classification because
reporting requirements are given in the regulations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
This request involves administrative changes only. The changes
consist of duplicates or overly burdensome reporting requirements or
the deletion of completed items required by [the TSs or] conditions
from the original issuance of Operating License NPF-30 [for the
Callaway Plant]. No actual plant equipment or accident analyses will
be affected by the proposed changes. Therefore, the proposed changes
do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This request involves administrative changes only. The changes
consist of duplicates or overly burdensome reporting requirements or
the deletion of completed items required by [the TSs or] conditions
from the original issuance of Operating License NPF-30. No actual
plant equipment or accident analyses will be affected by the
proposed change[s] and no failure modes not bounded by previously
evaluated accidents will be created. Therefore, the proposed changes
do not create a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel and fuel cladding, Reactor
Coolant System pressure boundary, and containment structure
[pressure boundary]) to limit the level of radiation dose to the
public. This request involves administrative changes only.
No actual plant equipment or accident analyses will be affected
by the proposed change[s]. The changes consist of duplicates or
overly burdensome reporting requirements or the deletion of
completed items required by [the TSs or] conditions from the
original issuance of Operating License NPF-30. Additionally, the
proposed changes will not relax any criteria used to establish
safety limits, will not relax any safety system settings, or will
not relax the bases for any limiting conditions of operation.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Robert A. Gramm.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: September 15, 2004.
Description of amendment request: The proposed changes will change
the Administrative Controls Section of the Technical Specifications
(TS) in order to incorporate title changes, change the location where
the plant-specific titles and TS titles are correlated, and relocate
the unit staff requirements to the Quality Assurance Program. These
proposed changes will support the implementation of proposed Virginia
Electric and Power Company Topical Report DOM-QA-1, ``Nuclear Facility
Quality Assurance Program Description,'' currently under NRC staff
review. In addition, these proposed TS changes eliminate the
descriptions of the onsite and offsite safety review organizations.
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. Operation of Surry Units 1 and 2 in accordance with the
proposed license amendments would not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change is administrative in nature and does not
affect plant systems, structures or components (SSCs) or plant
operation during normal or accident conditions. The proposed change
only affects the designated titles of personnel, rewords or
relocates requirements within TS or deletes requirements that are
either not required to be part of TS or are already required by
regulation. The change also relocates the detailed description of
the onsite and offsite safety review organizations and non-licensed
personnel qualification requirements to the Quality Assurance
Program. Therefore, this change has no bearing on the probability of
an accident. The management organizational structure and safety and
operational reviews have not changed and, therefore, do not impact
the ability of operating procedures or administrative controls to
prevent or mitigate
[[Page 70724]]
a previously evaluated accident. As such, this change does not alter
the conclusions of the existing safety analyses and therefore does
not alter the consequences of an accident previously evaluated.
2. Operation in accordance with the proposed license amendments
would not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed administrative change continues to ensure that
adequate management oversight exists at the plant in accordance with
the existing Technical Specifications. The proposed change only
affects the designated titles of personnel, rewords or relocates
requirements within TS or deletes requirements that are either not
required to be part of TS or are already required per regulation.
The change also relocates the detailed description of the onsite and
offsite safety review organizations and non-licensed personnel
qualification requirements to the Quality Assurance Program.
Therefore this change does not impact plant SSCs or plant operation
and therefore does not create the possibility of an accident of a
different type than evaluated previously. The management
organizational structure and safety and operational reviews have not
changed. Therefore, there is no change in the method of plant
operation, operation review or system design review. There are no
new or different accident scenarios, accident initiators, nor
failure mechanisms that will be introduced due to this change.
3. Operation in accordance with the proposed license amendments
would not involve a significant reduction in a margin of safety.
The proposed change only affects the designated titles of
personnel, rewords or relocates requirements within TS or deletes
requirements that are either not required to be part of TS or are
already required per regulation. The change also relocates the
detailed description of the onsite and offsite safety review
organizations and non-licensed personnel qualification requirements
to the Quality Assurance Program. Consequently, this change does not
impact plant design, plant operation or any safety margin and,
therefore, does not significantly reduce a margin of safety.
This evaluation concludes that the proposed amendments to the
Surry Units 1 and 2 Technical Specifications do not involve a
significant increase in the probability or consequences of a
previously evaluated accident, do not create the possibility of a
new or different kind of accident and 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: Ms. Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Section Chief: Mary Jane Ross-Lee (Acting).
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: July 23, 2004.
Description of amendment request: The amendment would revise
Technical Specification 3.6.3, ``Containment Isolation Valves,'' by (1)
Adding the abbreviation ``(CIV)'' for containment isolation valve in
Condition A of the Actions for the Limiting Condition for Operation;
(2) deleting the Note and revising Condition A to be for only one
penetration flow path with one CIV inoperable; (3) revising the
completion time for Required Condition A.1 from 4 hours to as much as 7
days depending on the category of the inoperable CIV; and (4) revising
Condition C to be for two or more penetration flow paths with one CIV
inoperable. The proposed amendment is based on Topical Report WCAP-
15791-P, ``Risk-Informed Evaluation of Extensions to Containment
Isolation Valve Completion Times.''
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 to the Completion Times do not
change the response of the plant to any accidents and have an
insignificant impact on the reliability of the containment isolation
valves. The containment isolation valves will remain highly reliable
and the proposed changes will not result in a significant increase
in the risk of plant operation. This is demonstrated by showing that
the impact on plant safety as measured by the large early release
frequency (LERF) and incremental conditional large early release
probabilities (ICLERP) is acceptable. These changes are consistent
with the acceptance criteria in [the risk-informed] Regulatory
Guides 1.174 and 1.177. Therefore, since the containment isolation
valves will continue to perform their [safety] functions with high
reliability as originally assumed and the increase in risk as
measured by LERF and ICLERP is acceptable, there will not be a
significant increase in the consequences of any accidents.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, or
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not alter or
prevent the ability of structures, systems, and components (SSCs)
from performing their intended [safety] function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed changes do not affect the source term,
containment isolation, or radiological release assumptions used in
evaluating the radiological consequences of an accident previously
evaluated. Further, the proposed changes do not increase the types
or amounts of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures. The proposed changes are consistent with the
safety analysis assumptions and resultant consequences [in Chapter
15, ``Accident Analysis,'' of the Updated Final Safety Analysis
Report (USAR) for the plant].
Therefore, it is concluded that this change does not increase
the probability of occurrence of a malfunction of equipment
important to safety.
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 do not result in a change in the manner in
which the containment isolation valves provide plant protection.
There are no design changes associated with the proposed changes.
The changes to Completion Times do not change any existing accident
scenarios, nor create any new or different accident scenarios.
The changes do 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. In addition,
the changes do not impose any new or different requirements or
eliminate any existing requirements. The changes do not alter
assumptions made in the safety analysis. The proposed changes are
consistent with the safety analysis assumptions and current plant
operating practice.
Therefore, the possibility of a new or different malfunction of
safety related equipment is not created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do 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 impacted by these changes. The proposed changes will not
result in plant operation in a configuration outside the design
basis. The calculated impact on risk is insignificant and is
consistent with the acceptance criteria contained in Regulatory
Guides 1.174 and 1.177.
Therefore, it is concluded that this 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
[[Page 70725]]
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Robert Gramm.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity For a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendments: October 29, 2004.
Brief description of amendments: Provide a one-time change to
Function 4a, ``Reactor Coolant System (RCS) Hot Leg Temperature
Indication,'' of Technical Specification Table 3.3.4-1. This would
allow continued operation until the next refueling outage (spring of
2005) with one out of four RCS hot leg temperature indications
inoperable in the Auxiliary Control Room.
Date of publication of individual notice in the Federal Register:
November 5, 2004 (69 FR 64596).
Expiration date of individual notice: November 19, 2004.
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, 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 NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, (301) 415-4737 or by e-mail to [email protected]. (Note: Public
access to ADAMS has been temporarily suspended so that security reviews
of publicly available documents may be performed and potentially
sensitive information removed. Please check the NRC Web site for
updates on the resumption of ADAMS access.)
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: August 27, 2004, as supplemented
by letters dated October 11 and 19, 2004.
Brief description of amendment: The amendment revised the Technical
Specifications, Section 2.1.A, changing the safety limit minimum
critical power ratio value from 1.09 to 1.10 for both four-or five-
recirculation-loop operation, and from 1.10 to 1.12 for three-
recirculation-loop operation.
Date of Issuance: November 16, 2004.
Effective date: November 16, 2004, and shall be implemented within
60 days of issuance.
Amendment No.: 252.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 14, 2004 (69
FR 55467). The October 11 and 19, 2004, letters provided clarifying
information within the scope of the original application and did not
change the staff's initial proposed no significant hazards
consideration determination. The Commission's related evaluation of
this amendment is contained in a Safety Evaluation dated November 16,
2004.
No significant hazards consideration comments received: No.
AmerGen Energy Company, LLC, et. al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station (OCNGS), Ocean County, New Jersey, Docket
No. 50-289, Three Mile Island Nuclear Station, Unit 1 (TMI-1), Dauphin
County, Pennsylvania
Date of application for amendments: March 23, 2004, as supplemented
June 16, 2004.
Brief description of amendments: The amendments relocate the
Independent Onsite Safety Review Group requirements from the
Administrative Controls in Section 6 of the Technical Specifications to
the Exelon Generation Company, LLC (EGC)/AmerGen Energy Company, LLC
(AmerGen) Quality Assurance Topical Report (QATR) at TMI-1 and OCNGS.
In addition, administrative corrections are included, which update
references to the EGC/AmerGen QATR, which has replaced the OCNGS and
TMI-1 Operational Quality Assurance Plans.
Date of issuance: November 8, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 251 and 252.
Facility Operating License Nos. DPR-16 and DPR-50: Amendments
revised the Technical Specifications.
Date of initial notices in Federal Register: May 11, 2004 (69 FR
26186).
The supplement dated June 16, 2004, 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 determinations.
The Commission's related evaluation of the amendments is contained
in a
[[Page 70726]]
Safety Evaluation dated November 8, 2004.
No significant hazards consideration comments received: No.
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendment: April 23, 2004.
Brief description of amendment: The amendment deletes Technical
Specification Section 6.16, ``Post-Accident Sampling Programs NUREG
0737 (II.B.3, II-F.1.2),'' and the related requirements to maintain a
Post Accident Sampling System.
Date of issuance: November 22, 2004.
Effective date: As of the date of issuance and shall be implemented
within 180 days.
Amendment No.: 253.
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 31, 2004 (69 FR
26187)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 22, 2004.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: March 12, 2004, and
supplemented by letters dated June 16 and September 2, 2004.
Brief description of amendments: The amendments modify the LaSalle
Technical Specifications (TS) to eliminate selected response time
testing requirements associated with Reactor Protection System
instrumentation and Primary Containment Isolation instrumentation for
Main Steam Line Isolation functions. Specifically, the changes revise
the response time testing requirements for TS Section 3.3.1.1,
``Reactor Protection System (RPS) Instrumentation,'' Reactor Vessel
Steam Dome Pressure--High function and TS Section 3.3.6.1, ``Primary
Containment Isolation Instrumentation,'' Reactor Vessel Water Level--
Low Low Low, Level 1 and Main Steam Line Pressure--Low functions.
Date of issuance: November 19, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 169, 155.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the TS.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19569).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 19, 2004.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of application for amendment: December 23, 2003.
Brief description of amendment: The amendment modifies technical
specification (TS) requirements to adopt the provisions of Industry/TS
Task Force (TSTF) change TSTF-359, ``Increased Flexibility in Mode
Restraints.''
Date of issuance: November 10, 2004.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 219.
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 16, 2004 (69
FR 55844).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 10, 2004.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: May 14, 2004.
Brief description of amendment: The amendment relocates the
requirements of Technical Specification 3.3(1)a, ``Reactor Coolant
System and Other Components Subject to ASME XI Boiler & Pressure Vessel
Code Inspection and Testing Surveillance'' and TS 3.4, ``Reactor
Coolant System Integrity Testing,'' to the Updated Safety Analysis
Report (USAR). Requirements in TS 3.3(1)a were related to inservice
inspection of ASME Class 1, 2, and 3 components and requirements in TS
3.4 were related to reactor coolant system integrity testing.
Date of issuance: November 8, 2004.
Effective date: November 8, 2004, and shall be implemented within
120 days from the date of issuance.
Amendment No.: 230.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: June 22, 2004 (69 FR
34703)
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated November 8, 2004.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: June 22, 2004, as supplemented
on September 27, 2004.
Brief description of amendments: The amendments revise the
frequency associated with Surveillance Requirement (SR) 3.3.8.1.4,
which directs the performance of the logic system functional test, from
once every 18 months to once every 24 months. The amendments change the
SRs in Hatch, Units 1 and 2 Technical Specifications.
Date of issuance: November 22, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 243/186.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Surveillance Requirements in the Technical
Specifications.
Date of initial notice in Federal Register: August 3, 2004 (69 FR
46592).
The supplement dated September 27, 2004, provided clarifying
information that did not change the scope of the June 22, 2004,
application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 22, 2004.
No significant hazards consideration comments received: No.
Tennessee Valley Authority (TVA), Docket No. 50-390, Watts Bar Nuclear
Plant, Unit 1, Rhea County, Tennessee
Date of amendment request: October 29, 2004, as supplemented
November 5, 2004.
Description of amendment request: The amendment provides a one-time
change to Function 4a, ``Reactor Coolant System (RCS) Hot Leg
Temperature Indication,'' of Technical Specification (TS) Table 3.3.4-1
to allow continued operations until the next refueling outage with one
out of four RCS Hot Leg Temperature Indications inoperable in the
Auxiliary Control Room.
[[Page 70727]]
Date of Issuance: November 19, 2004.
Effective date: As of the date of issuance and shall be implemented
immediately upon receipt.
Amendment No.: 53.
Facility Operating License No. (NPF-90): Amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. On November 5, 2004, the Commission issued a
notice (69 FR 64596) that included the staff's proposed determination
that the amendment request involves no significant hazards
consideration (NSHC). The notice provided an opportunity to submit
comments on the Commission's proposed NSHC determination. No comments
have been received. The notice also provided an opportunity to request
a hearing by November 19, 2004, but indicated that if the Commission
makes a final NSHC determination, any such hearing would take place
after issuance of the amendment. The supplement of November 5, 2004, is
within the scope of that notice, and did not change the proposed no
significant hazards consideration.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated November 19, 2004.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Michael L. Marshall, Jr.
Dated at Rockville, Maryland, this 29th day of November, 2004.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 04-26606 Filed 12-6-04; 8:45 am]
BILLING CODE 7590-01-P