[Federal Register Volume 68, Number 72 (Tuesday, April 15, 2003)]
[Notices]
[Pages 18269-18294]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-9026]
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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 Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from April 4, 2003, through April 17, 2003. The
last biweekly notice was published on April 1, 2003, (68 FR 15756).
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
[[Page 18270]]
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.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing 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
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By May 15, 2003, 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 2.714, which is
available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records will be accessible from the
Agencywide Documents Access and Management System's (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing
or petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, 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 factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, 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 with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of continuing disruptions in delivery of mail to United States
Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the request for
hearing and
[[Page 18271]]
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 because of continuing disruptions in delivery of mail
to United States Government offices, 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 filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
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. 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].
Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power
Plant, Unit 1 (Fermi 1), Monroe County, Michigan
Date of amendment request: January 28, 2003, (Reference NRC-03-
0011).
Description of amendment request: The proposed amendment will
revise the Technical Specifications by:
1. Section A.1, 2, 4, 8, C.1, D, E.1, H.3.b, I.5, I.7b, I.9.d have
been previously deleted and the word ``Deleted'' used as a place marker
to alleviate the need to renumber all sections. This request proposes
to remove these sections and renumber as appropriate.
2. Sections C.2 and E.2 cover the Reactor Building and Fuel and
Repair Building Drains. This request proposes to delete the
requirements in sections C.2 and E.2, which is all that remains in
sections C and E. Section C, Reactor Building, and E, Fuel and Repair
Building, will be deleted in their entirety.
3. Added, ``Monitoring or sampling for tritium will not be required
if the sample results have determined that tritium is not present
during a given evolution'' in Section F. This is to clarify the intent
of ``During other evolutions resulting in radioactive gaseous
effluents, the effluents shall be monitored or sampled and analyzed for
tritium and particulates.''
4. Section H.1 and 2 cover alarms, including surveillances, allowed
out of service time, compensatory measures and alarm readouts for
alarms associated with water intrusion. This request proposed to delete
these sections on water intrusion alarms.
5. Sections H.3 and 4 cover required inspections of the facility.
This request proposes to delete the requirement for radiation
surveillance of the steam cleaning room access plug, which is Item c.
of H.3, Fuel and Repair Building.
This proposal adds the words ``(until made inactive)'' to H.3
Reactor Building Item c. This request also proposes to delete recording
liquid waste tank levels, which is Item c. in Section H.4.
6. Table H-1 lists the required Fermi 1 alarms and their alarm
points. Only water intrusion alarms are currently covered in this
table. This request proposed to delete this alarm table.
7. Editorial changes are included in this proposed request. In
section I.2, the word ``employes'' will be changed to ``employees''. In
Section I.2.b the word ``He'' will be changed to ``The Health
Physicist''. In Section I.7 the word ``his'' will be removed from the
following sentence, ``The Custodian may temporarily change a procedure
by Written Order following his determination that the change does not
constitute a significant increase in the hazards associated with the
operation.'' In Section I.9.h the word ``usual'' will be changed to
``unusual''.
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 using the standards in 10 CFR 50.92(c). The licensee's
analysis is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident.
Removing the requirements for water intrusion monitoring, liquid
waste tanks level recording, and building drains will not
significantly increase the possibility of an accident as long as the
probability of an uncontrolled sodium and water reaction is not
significantly increased. This is accomplished by the amount of
volume of the area in which the sodium is present where water
intrusion is currently monitored. It would take a long period of
time for the water intrusion to reach the sodium piping and this
would still not increase the probability as long as the piping is
not breached. When the piping is breached during the sodium
abatement process, it will be completed under controlled conditions.
Removal of the instrumentation may delay the discovery of a liquid
spill but cannot affect the probability of the spill since it is
only instrumentation. The consequences of an accident will not be
increased because the previously analyzed accident accounts for all
of the radioactive material contained within the liquid waste tanks
and primary sodium to be released. This change will not increase the
amount of radioactive material. The editorial changes, steam
cleaning room plug radiation survey deletion, or the clarification
made to gaseous effluent monitoring for tritium will not
significantly increase the probability or consequences of an
accident, because they have no impact on how any systems are
operated or what systems are removed from the facility.
2. The proposed change does not create the possibility of a new
or different accident from any previously evaluated.
Removing the requirements for water intrusion monitoring and
liquid waste tanks level recording will not create the possibility
of a new or different accident from any previously evaluated. The
accidents these systems monitor for have already been analyzed for,
including a release of the radioactive sodium during a sodium and
water reaction and the release of the entire contents of the liquid
waste tanks. Removing the building drains requirements will not
cause a different type of accident since the drains only affect
where liquid flows. Where liquid flows cannot cause an accident
unless the drains place water where it does not belong. This can
only impact a liquid water release or sodium accident. The editorial
changes, survey deletion, and the clarification made to gaseous
effluent monitoring for tritium will not create the possibility of a
new or different accident, since they do not introduce any new modes
of operation of facility equipment.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The removal of the requirements for water intrusion monitoring,
liquid waste tanks level recording, and building drains may slightly
reduce the margin of safety, but not significantly. Removing them
does not in itself introduce water into the sodium containing
systems. Nor does removing them allow for an unmonitored discharge
of any radioactive effluents. Discharges are still controlled by
Section C of the proposed amendment to the Technical Specifications.
The decommissioning project is now ongoing and the facility no
longer normally vacant as it was during the initial time following
facility retirement. In addition, the calculated consequences of
releasing the radioactive material are small and within 10 CFR 20
limits. The editorial changes or survey deletion will not
significantly reduce a margin of safety, because the survey is of a
floor plug that has been removed from the entrance to an area and
has no function. The
[[Page 18272]]
clarification made to gaseous effluent monitoring for tritium will
not significantly reduce a margin of safety since tritium monitoring
is still required for evolutions involving sodium processing and
pipe cutting, and during other activities, unless results have
determined tritium is not present during a given evolution.
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, NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John Flynn, Esquire, Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Section Chief: Claudia M. Craig.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: November 25, 2002.
Description of amendment request: The amendments would revise the
Technical Specifications (TS) for the Ventilation Filter Testing
Program (VFTP), Annulus Ventilation System (AVS), Auxiliary Building
Filtered Ventilation Exhaust System (ABFVES), Fuel Handling Ventilation
Exhaust System (FHVES), and Control Room Area Ventilation System
(CRAVS), and containment penetrations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The following discussion is a summary of the evaluation of the
changes contained in this proposed amendment against the 10 CFR
50.92(c) requirements to demonstrate that all three standards are
satisfied. A no significant hazards consideration is indicated if
operation of the facility in accordance with the proposed amendment
would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated, 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.
First Standard
Does operation of the facility in accordance with the proposed
amendment involve a significant increase in the probability or
consequences of an accident previously evaluated?
This licensee amendment request proposes amendments to the
system TS and/or Bases and/or VFTP TS requirements for the AVS,
ABFVES, FHVES, and CRAVS. It also proposes amendments to the TS and
Bases for Containment Penetrations. The AVS is in standby during
normal plant operations and operates only following a Safety
Injection signal or during a test. It is not an accident initiator.
The ABFVES is in operation during normal plant operations. However,
the ABFVES is not used in direct support of any phase of power
generation or conversion or transmission, shutdown cooling, fuel
handling operations, or processing of radioactive fluids. Therefore,
it is not an accident initiator. The FHVES is utilized to support
fuel handling operations when moving recently irradiated fuel. It is
not an accident initiator. The CRAVS operates during normal plant
operations. However, it is not an accident initiator (the CRAVS
being defined so as to exclude equipment that maintains an
appropriately low temperature in the control room). The status of
containment penetrations is required to be controlled so as to
minimize the consequences of a fuel handling accident or a weir gate
drop accident. The containment penetrations by themselves are not
accident initiators. No accident initiators are associated with the
changes proposed in this license amendment request. For these
reasons, operation of the facility in accordance with this proposed
amendment does not involve a significant increase in the probability
of any accident previously evaluated.
In support of the proposed amendment, an analysis has been
performed to determine the radiological consequences of the design
basis LOCA [loss-of-coolant accident] at Catawba Nuclear Station.
The analysis made use of the Alternative Source Term (AST)
methodology and in general conformed to the regulatory positions of
Regulatory Guide 1.183, [``Alternative Radiological Source Terms for
Evaluating Design Basis Accidents at Nuclear Power Reactors,''
(ML003716792)
(Draft DG1081 Issued December 1999)] and the draft regulatory
positions of DG-1111. Total Effective Dose Equivalent (TEDE)
radiation doses at the Exclusion Area Boundary (EAB), boundary of
the Low Population Zone (LPZ), and to the control room operators
were calculated and found to be acceptable.
TEDE's have been estimated from the radiation doses with the
current analysis (reported in the UFSAR [Updated Final Safety
Analysis Report]) using the guidelines of Regulatory Guide 1.183
modified as reported in Appendix A of Attachment 3 [of the
licensee's submittal dated November 25, 2002]. These TEDE's are
compared to the limiting TEDE's from the proposed analysis as
follows:
TEDE's Following the Design Basis LOCA
------------------------------------------------------------------------
TEDE'S (Rem)
Location -----------------------
UFSAR Proposed
------------------------------------------------------------------------
EAB............................................. 9.95 7.21
LPZ............................................. 1.90 3.97
Control Room.................................... 1.57 2.65
------------------------------------------------------------------------
The new value for the control room TEDE radiation dose is higher
than the TEDE radiation dose equivalent to the radiation doses
currently reported in the UFSAR. However, the limiting control room
TEDE radiation dose reported in this submittal is lower than the
acceptance criterion by 47%. The new LPZ TEDE radiation dose is
higher than the equivalent TEDE radiation dose currently
represented. On the other hand, the margin to the acceptance
criterion is 84%. The TEDE radiation doses newly computed at the EAB
for the design basis LOCA is lower than the corresponding equivalent
EAB TEDE radiation dose currently represented in the UFSAR. The
margin in the EAB TEDE radiation dose to the guideline value is 71%.
In all cases, there is significant margin between the newly
calculated post-LOCA TEDE radiation doses and the corresponding
regulatory guideline values. In the sense that the margins to the
germane regulatory guideline values are still large, the new values
of TEDE radiation doses are comparable to the equivalent TEDE
associated with the post-LOCA radiation doses currently listed in
the UFSAR. Therefore, the proposed amendment is determined to not
result in a significant increase in accident consequences.
The changes proposed to the TS for Containment Penetrations are
editorial in nature and will have no effect upon accident
consequences.
The changes proposed to the VFTP TS for the AVS, ABFVES, and
FHVES will not result in a significant increase in any accident
consequences. The changes to make the penetration values for Unit 2
consistent with Unit 1 for the AVS, ABFVES, and FHVES are acceptable
because the appropriate safety factors as delineated in the
applicable regulatory guideline documents are still maintained. The
change to the flowrate specified for the ABFVES is consistent with
the design basis operation of this system. Also, the editorial
changes proposed to the VFTP TS will have no impact on any
accidents.
Operation of the facility in accordance with the proposed
amendment does not involve a significant increase in the
consequences of an accident previously evaluated.
Second Standard
Does operation of the facility in accordance with the proposed
amendment create the possibility of a new or different kind of
accident from any accident previously evaluated?
This proposed amendment does not involve addition, removal, or
modification of any plant system, structure, or component. These
changes will not affect the operation of any plant system,
structure, or components as directed in plant procedures.
[[Page 18273]]
The analysis performed in support of this license amendment
request, together with the analyses of the design basis fuel
handling accident and weir gate drop reported in previously
submitted and NRC approved license amendment requests, includes full
scope implementation of AST methodology. This analysis does not
represent any change in the post-accident operation of any plant
system, structure, or component.
Operation of the facility in accordance with this amendment does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
Third Standard
Does operation of the facility in accordance with the proposed
amendment involve a significant reduction in the margin of safety?
Margin of safety is related to confidence in the ability of
fission product barriers to perform their design functions following
any of their design basis accidents. These barriers include the fuel
cladding, the Reactor Coolant System, and the containment. The
performance of these barriers either during normal plant operations
or following an accident will not be affected by the changes
associated with the license amendment request.
The AVS is associated with the containment fission product
barrier. Its post-accident operation will not be affected by
implementation of the amendment to its TS. The operation of the
ABFVES either during normal plant operations or following an
accident will not be affected by implementation of the amendment to
its TS. The operation of the FHVES either during normal plant
operations or following an accident will not be affected by
implementation of the amendment to its TS. The operation of the
CRAVS either during normal plant operations or following an accident
will not be adversely affected by the proposed changes to its TS
Bases. The operation of Containment Penetrations following an
accident will not be adversely affected by the proposed change to
its TS.
As noted, an analysis of radiological consequences of the design
LOCA at Catawba Nuclear Station has been performed in support of
this license amendment request. The design basis LOCA scenarios were
selected based on extensive evaluations of Catawba, its design
basis, and its anticipated response to a design basis LOCA. Credit
was taken only for safety related systems, structures, and
components in simulating the mitigation of radiological consequences
of the LOCA. Limiting values were taken for performance
characteristics of the Class 1E systems modeled in the analysis. The
radiological consequences (TEDE radiation doses at the EAB, LPZ, and
in the control room) are within the regulatory guideline values with
significant margin.
The changes proposed to the VFTP TS for the AVS, ABFVES, and
FHVES will not result in a significant reduction in the margin of
safety. These changes are supported by regulatory guidance
documents, and are consistent with existing system operation. Also,
the editorial changes proposed to the VFTP TS will not have any
impact on safety.
Operation of the facility in accordance with the proposed
amendment does not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, and Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, located in Mecklenburg County, North
Carolina and York County, South Carolina
Date of amendment request: November 20, 2002, as supplemented
January 21, 2003.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) for REQUIRED ACTIONS requiring
suspension of operations involving positive reactivity additions and
various NOTES that preclude reduction in boron concentration. The
proposed changes revise these REQUIRED ACTIONS and NOTES to limit the
introduction of positive reactivity such that the required margin to
criticality, the shutdown margin and refueling boron concentration
limits will still be satisfied. The licensee stated that the changes
are consistent with the Technical Specification Task Force (TSTF)
traveler number 286, Revision 2. Associated changes are also proposed
for the TS Bases. Basis for proposed no significant hazards
consideration determination: As required by 10 CFR 50.91(a), the
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
The following discussion is a summary of the evaluation of the
change contained in this proposed amendment against the 10 CFR 50.92
(c) requirements to demonstrate that all three standards are
satisfied. A ``no significant hazards consideration'' is indicated
if operation of the facility in accordance with the proposed
amendment would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated, 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.
First Standard
The proposed changes do not involve any physical alteration of
plant systems, structures, or components. The proposed changes
revise ACTIONS in the Catawba Nuclear Station (CNS) and McGuire
Nuclear Station (MNS) Technical Specifications (TS) that require
suspending operations involving positive reactivity additions and
several Limiting Condition for Operation (LCO) Notes that preclude
reduction in boron concentration. The change revises these ACTIONS
and LCO Notes to limit the introduction of reactivity such that the
required SHUTDOWN MARGIN (SDM) or refueling boron concentration will
still be satisfied. The proposed change ensures that the reactivity
condition [keff] specified in mode definition, the SDM of
LCO 3.1.1 and minimum boron concentration requirements of LCO 3.9.1
are met. Therefore, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated in the updated final safety analysis
report (UFSAR) because the accident analysis assumptions and initial
conditions will continue to be maintained.
Second Standard
The proposed changes do not involve any physical alteration of
plant systems, structures, or components. The proposed changes,
which allow positive reactivity additions that do not result in the
SDM or the refueling boron concentration being exceeded, do not
introduce new failure mechanisms for system structures, or
components not already considered in the UFSAR. Therefore, the
possibility of a new or different kind of accident from any accident
previously evaluated is not created because no new failure
mechanisms or initiating events have been introduced.
Third Standard
The proposed changes do not involve a significant reduction in a
margin of safety because the ability to make the reactor subcritical
and maintain it subcritical during all operating conditions and
modes of operation will be maintained. The margin of safety is
defined by the SDM of LCO 3.1.1 and minimum boron concentration
requirements of LCO 3.9.1. The proposed changes do not affect these
operating restrictions and the margin of safety, which assures the
ability to make and maintain the reactor subcritical, is not
affected.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
[[Page 18274]]
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, and Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, located in Mecklenburg County, North
Carolina and York County, South Carolina
Date of amendment request: January 31, 2003.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) to incorporate an asymmetrical
ice mass distribution within the ice condenser containment (ICC) by
specifying revised safety analysis ice mass quantity requirements for
three specific radial zones of the ice bed. Associated changes to the
Bases were also proposed.
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:
Duke Energy Corporation (Duke) has concluded that operation of
Catawba Nuclear Station (CNS) Units 1 & 2, and McGuire Nuclear
Station (MNS) Units 1 & 2, in accordance with the proposed changes
to the Technical Specifications (TS) does not involve a significant
hazards consideration. Duke's conclusion is based on its evaluation,
in accordance with 10 CFR 50.91(a)(1), of the three standards set
forth in 10 CFR 50.92(c).
A. The Proposed Change Does Not Involve a Significant Increase
In The Probability or Consequences Of An Accident Previously
Evaluated.
The only analyzed accidents of possible consideration in regards
to changes potentially affecting the ice condenser are a loss of
coolant accident (LOCA) and a high energy line break (HELB) inside
containment. However, the ice condenser is not postulated as being
the initiator of any LOCA or HELB. That is because it is designed to
remain functional following a design basis earthquake, and the ice
condenser does not interconnect or interact with any systems that
interconnect or interact with the Reactor Coolant or Main Steam
Systems. Since these proposed changes do not result in, or require,
any physical change to the ice condenser that could introduce an
interaction with the Reactor Coolant or Main Steam Systems, then
there can be no change in the probability of an accident previously
evaluated.
Regarding consequences of analyzed accidents, the ice condenser
is an engineered safety feature designed, in part, to limit the
containment sub-compartment and containment vessel pressure
immediately following the initiation of a LOCA or HELB. Conservative
sub-compartment and containment pressure analysis [based on the
proposed changes] shows these criteria will be met if the total ice
mass within the ice bed is maintained in accordance with the DBA
[Design Basis Accident] analysis; therefore, the proposed TS SR
[Surveillance Requirement] changes of these requirements will not
increase the consequences of any accident previously evaluated.
Thus, based on the above, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
B. The Proposed Change Does Not Create The Possibility Of A New
Or Different Kind Of Accident From Any Accident Previously
Evaluated.
As previously described, the ice condenser is not postulated as
being the initiator of any design basis accident. The proposed
changes do not impact any plant system, structure or component that
is an accident initiator. The proposed TSs and TS Bases changes do
not involve any hardware changes to the ice condenser or other
change that could create any new accident mechanisms. Therefore,
there can be no new or different accidents created from those
already identified and evaluated.
C. The Proposed Change Does Not Involve A Significant Reduction
In A Margin Of Safety.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their design functions
during and following an accident situation. These barriers include
the fuel cladding, the reactor coolant system, and the containment
system. The performance of the fuel cladding and the reactor coolant
system will not be impacted by the proposed changes. The Application
provides a description of additional sub-compartment and containment
pressure response analysis that has been performed. This analysis
demonstrates that containment will remain fully capable of
performing its design function with implementation of the proposed
changes. Therefore, no safety margin will be significantly impacted.
Ice Condenser plant historical operating experience has shown
that the condition of the ice condenser can be ensured to be fully
capable of performing its specified safety functions with performing
ice mass verifications and ice mass distribution SRs on an 18 month
frequency. The request to increase the MNS [McGuire] surveillance
interval from 9 months to 18 months will provide performance of ice
mass verification at the end of the fuel cycle, which will verify
that the maintenance program is effective in maintaining the ice
mass for the entire fuel cycle. Duke's utilization of the data from
previous performance of TS required ice mass inspections, and
additional inspection beyond these requirements, has enabled the
development of a maintenance program that is reliably predictive
regarding the specific operating characteristics of each [of] the
ice beds at Catawba and McGuire Nuclear Stations. This maintenance
program reliably predicts sublimation and determines which ice
baskets to replenish prior to beginning a new 18 months operating
cycle. An ice mass surveillance performed at the conclusion of the
18 month frequency in an as-found condition verifies that the
maintenance program is restoring the ice bed operating cycle to
maintain the ice mass quantity and distribution requirements for
performance of the intended safety functions.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
Duke Energy Corporation, Docket No. 50-370, McGuire Nuclear Station,
Unit 2, Mecklenburg County, North Carolina
Date of amendment request: January 31, 2003.
Description of amendment request: The proposed amendment would
authorize the licensee to change the Updated Final Safety Analysis
Report (UFSAR) to describe a process for the intentional puncture of an
irradiated fuel rod in order to transfer the fuel rod gap gasses to a
collection chamber, and then straighten the fuel rod for storage in a
broken rod capsule.
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:
Duke Energy has evaluated whether or not a significant hazards
consideration is involved with the proposed amendment by focusing on
the three standards set forth in 10 CFR 50.92, ``Issuance of
amendment,'' as discussed below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The bent rod, located in the McGuire Unit 2 spent fuel pool, has
no interfaces with any primary system, secondary system, or power
transmission system. All work will be performed in the spent fuel
pool, with the bent rod located under approximately 23 feet of
water. None of the systems listed above are modified by the
activity. No accident initiator or accident mitigation systems, for
any UFSAR [Updated Final Safety Analysis Report] Chapter 15
accidents, other than fuel handling accidents, are affected with
this proposed procedure for degassing and straightening of the
irradiated Mk-BW fuel rod. For these reasons, the activity does not
involve an increase in the probability of an accident previously
evaluated.
This evolution is bounded by the UFSAR Chapter 15 dropped fuel
assembly fuel handling accident inside the fuel handling building.
This accident assumes that the postulated accident occurs 100 hours
after reactor shutdown, the fuel assembly had 60 GWD/MTU [Gigawatt
Days/Metric Ton Uranium] burnup, all rods in one fuel
[[Page 18275]]
assembly are ruptured, and the assembly damaged has the highest
peaking factor. The resultant Exclusion Area Boundary doses for the
UFSAR Chapter 15 accident are 0.8 Rem Whole Body and 9.1 Rem
Thyroid.
For the planned evolution, the cladding on only one rod will be
breached and the fission product gas contained. This evolution will
occur approximately ten years after reactor shutdown. The fuel rod
burnup is only 20.46 GWD/MTU, and the fuel pin peaking factor is
1.28. Some accident mitigation will be provided by the fuel building
ventilation system filters, although the majority of the activity
will be from Kr-85, a noble gas, which is unaffected by these
filters. The highest potential dose occurs to a worker in the fuel
building, with whole body doses of less than 3 mRem and a thyroid
dose of less than 3E-11 mRem. Doses at the Exclusion Area Boundary
are trivial.
Should the gas container fail, the offsite activity release and,
as such, the consequences of this accident will be less than any
previously evaluated. Analyses have been performed to determine
upper bounds for the source term, the offsite doses, and the control
room dose. Both the source term and doses were found to be
significantly lower than the results of the corresponding design
basis analyses.
For the above reasons, it is determined that the intentional
degassing of the Mk-BW fuel rod does not involve a significant
increase in either the probability or the 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?
As discussed above, no ``accident initiators'' are affected by
the proposed activity. The planned evolution is bounded by the
dropped fuel assembly fuel handling accident inside the fuel
handling building. The fuel rod straightening and degassing tools
are no heavier than other fuel handling tools utilized in the spent
fuel pool during routine operations. A safety tray will be placed on
top of the racks and below the work area to capture any falling
debris during the operation. Also a mockup operation will be
performed at the Framatome facilities to identify and correct any
deficiencies in the tools and processes.
For these reasons, the activity will not create the possibility
of a new or different type of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Margin of safety is associated with the confidence in the
ability of the fission product barriers (the fuel and fuel cladding,
the reactor coolant system pressure boundary, and the containment)
to limit the level of radiation doses to the public. The proposed
degassing of the fuel rod will intentionally breach the fuel rod
cladding, but the fuel rod gap gasses will be captured in a
collection chamber for holdup and later controlled release.
This evolution will occur beyond a nine year cooling and
isotopic decay period. The level of activity in the fuel rod is very
low compared to the level of activity associated with the postulated
fuel handling accident; the only significant activity remaining is
approximately 10 Ci [Curies] of Krypton 85. The bent rod will be
maintained under 23 feet of water. Should the collection chamber
fail, and the fuel rod gap gas activity released, the highest
potential dose occurs to a worker in the fuel handling building,
with whole body doses of less than 3 mRem, and a thyroid dose of
less than 3E-11 mRem. For this reason, the resulting dose to the
public is inconsequential. Both offsite doses and doses to the
control room were found to be small compared to the limits of 10 CFR
100 and GDC 19. For these reasons, the activity does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: March 14, 2003.
Description of amendment request: The licensee requests
modification of the River Bend Technical Specifications to revise
several of the Surveillance Requirements (SRs) pertaining to testing of
the Division 1 and 2 standby diesel generators (DGs). The proposed
change would modify specific restrictions associated with these SRs
that prohibit performing required testing in Modes 1 and 2. The
affected SRs are SR 3.8.1.9 and SR 3.8.1.10.
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 DG and its associated emergency loads are accident
mitigating features, not accident initiating equipment. Therefore,
there will be no impact on any accident probabilities by the
approval of the requested amendment.
The design of plant equipment is not being modified by these
proposed changes. As such, the ability of the DG to respond to a
design basis accident will not be adversely impacted by these
proposed changes. The capability of the DG to supply power in a
timely manner will not be compromised by permitting performance of
DG testing during periods of power operation. Additionally, limiting
testing to only one DG at a time ensures that design basis
requirements for backup power is met, should a fault occur on the
tested DG. Therefore, there would be no significant impact on any
accident consequences.
Based on the above, the proposed change to permit certain DG
surveillance tests to be performed during plant operation will have
no effect on accident probabilities or consequences. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident causal mechanisms would be created as a result
of NRC [U.S. Nuclear Regulatory Commission] approval of this
amendment request since no changes are being made to the plant that
would introduce any new accident causal mechanisms. Equipment will
be operated in the same configuration with the exception of the
plant mode in which the testing is conducted. This amendment request
does not impact any plant systems that are accident initiators;
neither does it adversely impact any accident mitigating systems.
Based on the above, implementation of the proposed changes would
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their design functions
during and following an accident situation. These barriers include
the fuel cladding, the reactor coolant system, and the containment
system. The proposed changes to the testing requirements for the DG
do not affect the operability requirements for the DG, as
verification of such operability will continue to be performed as
required. Continued verification of operability supports the
capability of the DG to perform its required function of providing
emergency power to plant equipment that supports or constitutes the
fission product barriers.
Consequently, the performance of these fission product barriers
will not be impacted by implementation of this proposed amendment.
In addition, the proposed changes involve no changes to
setpoints or limits established or assumed by the accident analysis.
On this and the above basis, no safety margins will be impacted.
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 18276]]
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 Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.
Energy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: February 27, 2003.
Description of amendment request: The proposed amendment deletes
requirements from the technical specifications (TS) and other elements
of the licensing bases to maintain a post accident sampling system
(PASS). Licensees were generally required to implement PASS upgrades as
described in NUREG-0737, ``Clarification of TMI [Three Mile Island]
Action Plan Requirements,'' and Regulatory Guide 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI Unit 2. Requirements related to
PASS were imposed by Order for many facilities and were added to or
included in the TS for nuclear power reactors currently licensed to
operate. Lessons learned and improvements implemented over the last 20
years have shown that the information obtained from PASS can be readily
obtained through other means or is of little use in the assessment and
mitigation of accident conditions.
The changes are based on Nuclear Regulatory Commission (NRC)-
approved Technical Specification Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-413, ``Elimination of Requirements
for a Post Accident Sampling System (PASS).'' The NRC staff issued a
notice of opportunity for comment in the Federal Register (FR) on
December 27, 2001 (66 FR 66949), on possible amendments concerning
TSTF-413, 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 FR on March 20, 2002 (67 FR 13027). The licensee
affirmed the applicability of the following NSHC determination in its
application dated February 27, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident from any Previously Evaluated
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radioisotopes
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this 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: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Section Chief: Richard J. Laufer.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: March 20, 2003.
Description of amendment request: This proposed change reflects an
expanded operating domain for Vermont Yankee Nuclear Power Station (VY)
resulting from the proposed implementation of the Average Power Range
Monitor, Rod Block Monitor Technical Specifications/Maximum Extended
Load Line Limit Analysis (ARTS/MELLLA).
[[Page 18277]]
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR) 50.91(a), the licensee has provided its analysis
of the issue of no significant hazards consideration. The NRC staff has
reviewed the licensee's analysis against the standards of 10 CFR
50.92(c). The NRC staff's review is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change involves allowing VY to operate in an
expanded operating domain. Physical changes provide for enhanced
instrument performance or were the result of safety analyses that
support mitigation of design bases accidents. There are no changes
to radioactive source terms or release pathways. The proposed change
does not result in any significant change in the availability of
logic systems or safety-related systems themselves. Required
protective functions will be maintained. The proposed change does
not degrade plant design, operation, or the performance of any
safety system assumed to function in the accident analysis.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Create the possibility for a new or different kind of
accident from any previously evaluated.
The proposed change, which allows VY to operate in an expanded
operating domain, does not introduce any new accidents or failure
mechanisms because the change and the effects on existing
structures, systems and components have been evaluated and found to
not have any adverse effects. The proposed change will not
substantially impose new requirements or eliminate any existing
requirements.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident than those previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change, which allows VY to operate in an expanded
operating domain, does not alter the manner in which safety limits,
limiting safety system settings, or limiting conditions for
operation are determined. There is no impact on the conclusions of
any safety analysis. The proposed change does not involve any
increase in calculated off-site dose consequences. The performance
of equipment will not be significantly affected.
Therefore, there is no significant reduction in the margin of
safety as a result of this proposed change.
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. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: January 31, 2003.
Description of amendment request: The proposed amendments would
change the Technical Specifications (TS) allowable values (AVs) for
isolation condenser system isolation Function 4.a, Steam Flow-High, and
Function 4.b, Return Flow-High.
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 TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed TS changes support the replacement of a
differential pressure switch with a functionally equivalent
differential pressure switch. Since there are no functional changes
and no change in analytical limits, there is no significant increase
in the probability or consequences of an accident previously
evaluated.
Additionally, these changes will not increase the consequences
of an accident previously evaluated because the proposed changes do
not adversely impact structures, systems, or components.
Furthermore, there will be no change in the types or significant
increase in the amounts of any effluents released offsite as a
result of the proposed change.
In summary, the proposed changes do not involve a significant
increase in the probability of consequences of an accident
previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The change does not adversely impact the manner in which the
instrument will operate under normal and abnormal operating
conditions. Therefore, these changes provide an equivalent level of
safety and will not create the possibility of a new or different
kind of accident from any accident previously evaluated. The changes
in allowed values do not affect the current safety analysis
assumptions. Therefore, these changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed changes do not affect the probability of failure or
availability of the affected instrumentation. The revised AVs do not
affect the analytical limits assumed in the safety analyses for
actuation of instrumentation. Therefore, the proposed changes do not
result in a reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: February 17, 2003.
Description of amendment request: The proposed amendment would
revise Technical Specification (ITS) 3.6.3 ``Containment Isolation
Valves,'' to allow verification by administrative means of isolation
devices in high radiation areas, and isolation devices that are locked,
sealed or otherwise secured. The specific Conditions and Surveillance
Requirements (SR) in ITS 3.6.3 that will be affected are: (1) Condition
A--Required Action A.2, (2) Condition B--Required Actions B.1 and B.2,
(3) Condition C--Required Action C.2, and (4) SR 3.6.3.3 and SR
3.6.3.4. The licensee stated that the changes are consistent with the
NUREG-1430, ``Standard Technical Specifications: Babcock and Wilcox
Plants,'' Revision 2, and Standard Technical Specification Task Force
(TSTF) Traveler TSTF-440. Associated changes are also proposed for the
ITS Bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does not involve a significant increase in the probability or
consequences of an accident previously analyzed.
The proposed License Amendment Request (LAR) will revise the
position verification requirements for manual containment isolation
devices that are locked, sealed, or otherwise secured in the closed
position. The proposed changes will allow the use of administrative
controls to verify the position of these types of devices when they
are being used to meet the Required Actions of ITS 3.6.3 Condition
A, Condition B or Condition C, and will exclude these valves from
Surveillance Requirement (SR) 3.6.3.3 and
[[Page 18278]]
SR 3.6.3.4 physical position verification requirements.
The design function of the affected containment isolation
valves, and the initial conditions for accidents that require these
valves to be closed, will not be affected by the proposed changes.
Therefore, the changes will not increase the probability or
consequences of an accident previously evaluated.
2. Does not create the possibility of a new or different kind of
accident from any accident previously analyzed.
The proposed license amendment will revise the position
verification requirements for manual containment isolation devices
that are locked, sealed, or otherwise secured in the closed
position.
No changes to the actual position/status of these valves are
proposed by this amendment. The proposed amendment will not result
in changes to the design, physical configuration or operation of the
plant. Therefore, the proposed change will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does not involve a significant reduction in the margin of
safety.
Changes to the position verification requirements of normally
closed manual containment isolation valves that are locked, sealed,
or otherwise secured do not change the position/status of these
valves. The proposed amendment does not impact the ability of these
valves to perform their design function of controlling containment
leakage rates during design basis radiological accidents. Therefore,
the proposed amendment does not result in a reduction of 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602-1551.
NRC Section Chief: Allen G. Howe.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: March 11, 2003.
Description of amendment request: The proposed amendment would
change the operating license to authorize the licensee to revise the
updated final safety analysis report (UFSAR) by deleting a footnote
stating that the Nuclear Regulatory Commission (NRC) does not endorse
the reactor building crane as single-failure-proof.
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 will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
For heavy load handling associated with the spent fuel pool,
Section 5.1.4(2) of NUREG-0612 states ``The effects of heavy load
drops in the reactor building should be analyzed to show that the
evaluation criteria of Section 5.1 are satisfied.''
An alternative to this is Section 5.1.4(1): ``The reactor
building crane, and associated lifting devices used for handling of
* * * heavy loads, should satisfy the single-failure-proof
guidelines of Section 5.1.6 of this report.''
The upgraded crane and handling systems satisfy the guidelines
of Section 5.1.6. The evaluation criteria of NUREG-0612, Section 5.1
are met with a single-failure-proof crane that satisfies the
guidelines of Section 5.1.6, or consequence analysis that satisfies
Section 5.1.4(2).
Section 5.2 of NUREG-0612 states that an evaluation of fault
trees shows that: ``(1) The likelihood for unacceptable consequences
in terms of excessive releases of gap activity or potential for
criticality due to accidental dropping of postulated heavy loads
after implementation of the guidelines of Section 5.1 is very low;
and (2) The potential for unacceptable consequences is comparable
for any of the alternatives evaluated for fault trees, indicating
the relative equivalency between alternatives.''
Since the NRC fault tree evaluation shows that the potential for
unacceptable consequences is comparable for the two alternatives in
Section 5.1.4 of NUREG-0612, the proposed request does not
significantly change the potential for unacceptable consequences to
the plant in conducting heavy load handling above the spent fuel
pool. The probability of a load drop accident caused by use of the
reactor building crane has been reduced to where it is so small to
be considered not credible within regulatory accepted standards. The
reason for this is attributed to the following:
(a) The reactor building crane is single-failure-proof. In 1985,
the DAEC [Duane Arnold Energy Center] Reactor Building Crane was
modified to meet the requirements of NUREG-0554 ``Single Failure
Proof Cranes for Nuclear Power Plants.'' The design of the Ederer
hoist and trolley system was evaluated in a Staff SER [Safety
Evaluation Report] of the Generic Licensing Topical Report EDR-1,
Rev. 3, for Ederer's Nuclear Safety-Related Extra Safety and
Monitoring (X-SAM) Cranes, dated August 3, 1983.
(b) The rigging used with the crane will be single-failure-proof
per Section 5.1.6 of NUREG-0612.
(c) The requirements of NUREG-0612 Phase 1 have been
implemented. The NRC provided a Safety Evaluation (SE) and Technical
Evaluation Report (TER) by letter dated June 12, 1984 that concluded
that the guidelines of NUREG-0612, Sections 5.1.1 and 5.3 had been
satisfied and that Phase I of this issue for the DAEC was
acceptable.
Therefore, this proposed change will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The crane has been upgraded to meet single-failure-proof
requirements in accordance with the applicable provisions of NUREG-
0612 and NUREG-0554. The use of a single-failure-proof crane with
rigging and procedures that implement the requirements of NUREG-0612
assures that a cask drop is not credible. The loading on the single-
failure-proof crane will not exceed the design rated load of the
crane.
Rigging for critical loads will meet NUREG-0612 requirements for
single-failure-proof handling systems whenever a critical load is to
be lifted over safety related equipment, or over the spent fuel
pool, or over the cask when it is in the reactor building and loaded
with fuel. When a cask is loaded on the crane hook, the crane
trolley and bridge movements will be maintained within well defined
limits of operation.
The loading conditions, load combinations, allowable stress
limits, and methods of analysis used in the evaluations are
consistent with the current licensing basis for the DAEC and NRC
approved methods.
Therefore, this proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment will not involve a significant
reduction in a margin of safety.
In 1985, the reactor building crane was upgraded to single-
failure-proof in compliance with NUREG-0554. The upgraded crane and
handling system is in compliance with NUREG-0612, Sections 5.1.1 and
5.1.6. The NRC in NUREG-0612, Section 5.2 documented their review of
the potential consequences of a load drop when handled by a single-
failure-proof crane using single-failure-proof rigging compared with
other alternatives and concluded as follows: ``The likelihood for
unacceptable consequences in terms of excessive releases of gap
activity or potential for criticality due to accidental dropping of
postulated heavy loads after implementation of the guidelines of
Section 5.1 is very low.''
This means that a load drop is considered to be unlikely within
regulatory accepted standards when the load is handled by a single-
failure-proof crane and handling system, and performed in accordance
with Section 5.1 of NUREG-0612. A single-failure-proof crane design
incorporates the applicable design basis event that in this case is
a seismic event. A load drop is of such low probability that it is
considered unlikely when it is handled with the reactor building
crane since the crane and its handling systems satisfy the NUREG-
0612 criteria for a single-failure-proof crane. Therefore, any load
lifted over the spent fuel pool using the reactor building crane has
a very low probability of falling into the spent fuel pool
[[Page 18279]]
accidentally or as a result of a design basis event.
Therefore, this proposed amendment will 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. Alvin Gutterman, Morgan Lewis, 1111
Pennsylvania Avenue NW., Washington, DC 20004.
NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: January 29, 2003.
Description of amendment request: The proposed amendment would
change the drywell leakage and sump monitoring detection section of the
Technical Specifications (TSs). These proposed changes clarify the
definitions and restructure the coolant leakage section of the TSs and
revise unidentified leakage and total leakage requirements. The
revisions add a TS Limiting Condition for Operation for leakage-
detection instrumentation being inoperable. This request supercedes the
Nuclear Management Company's license amendment request of October 8,
2002, as supplemented November 8, 2002, which was previously noticed in
the Federal Register on October 17, 2002 (67 FR 64144).
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 will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed Technical Specification changes do not introduce
new equipment or new equipment operating modes, nor do the proposed
changes alter existing system relationships. Additionally, the
proposed changes do not affect any accident previously evaluated in
the Monticello Updated Safety Analysis Report (USAR). The changes
simply redefine the parameters for evaluation of leakage in the
drywell. The evaluation criteria for drywell leakage have been
refocused into the areas that are most susceptible to IGSCC
[intergranular stress corrosion cracking]. Consequently, the
probability of an accident previously evaluated is not significantly
increased.
The equipment referenced in the proposed changes is still
required to monitor the reactor coolant system operational leakage
to ensure appropriate action is taken before the integrity of the
reactor coolant pressure boundary is impaired. As a result,
operation of the facility with the proposed changes will continue to
meet the licensing basis and applicable guidelines. As such, the
consequences of any accident previously evaluated are not
significantly affected.
Therefore, the proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
The proposed changes do not involve physical alterations of the
plant; no new or different type of equipment will be installed; nor
are there significant changes in the methods governing normal plant
operation. The changes simply redefine the parameters for evaluation
of leakage in the drywell. The evaluation criteria for drywell
leakage have been refocused into the areas that are most susceptible
to IGSCC. Additionally, the changes do not create any new failure
mechanisms, malfunctions, or accident initiators not already
considered in the design and licensing bases.
Therefore, the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously analyzed.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
The proposed amendment redefines the parameters for evaluation
of leakage in the drywell. There are no physical alterations of the
plant; no new or different type of equipment will be installed; nor
are there significant changes in the methods governing normal plant
operation. Additionally, the proposed changes do not exceed or alter
a design basis or safety limit as established in the Monticello
licensing basis.
Therefore, these proposed changes will 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: Jay E. Silberg, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: February 11, 2003.
Description of amendment request: The proposed amendments would
revise technical specification (TS) 5.5.9, ``Ventilation Filter Testing
Program (VFTP)'' by (1) incorporating filter test face velocity limits
for the control room special ventilation system, auxiliary building
special ventilation system, spent fuel pool special and inservice purge
ventilation system, and shield building ventilation system; and (2)
making editorial changes. The proposed amendments would also delete the
additional conditions in Appendix B of the Operating Licenses which
require the licensee to complete an evaluation of the maximum test face
velocity for the ventilation systems in TS 5.5.9. The additional
conditions also require the licensee to submit a license amendment
request for a TS amendment to specify the maximum test face velocity if
the maximum actual face velocity is the greater than 110 percent of 40
fpm. Additionally, the proposed amendments would revise the penetration
and system bypass limit from 0.05 percent to 0.5 percent for the
ventilation systems.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Revision of the Allowable Filtration Penetration and System Bypass
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
This license amendment request proposes to increase the
penetration and system bypass limit for the control room special
ventilation system, auxiliary building special ventilation system,
spent fuel pool special and inservice purge ventilation system and
shield building ventilation system from 0.05% to 0.5%. These
ventilation systems are included in the plant design to mitigate
accident consequences and are not assumed accident initiators, thus,
this change does not involve a significant increase in the
probability of an accident. This change will assure that the subject
ventilation systems will perform within their intended design ranges
thus, this change assures that the consequences of an accident are
not increased.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
This proposed change does not alter the design, function, or
operation of any plant component and does not install any new or
different equipment. The malfunction of safety related equipment,
assumed to be operable in the accident analyses, would not be caused
as a result of the proposed Technical Specification change. No new
[[Page 18280]]
failure mode has been created and no new equipment performance
burdens are imposed. Therefore the possibility of a new or different
kind of accident from those previously analyzed has not been
created.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
This license amendment request proposes to increase the
penetration and system bypass limit for the control room special
ventilation system, auxiliary building special ventilation system,
spent fuel pool special and inservice purge ventilation system and
shield building ventilation system from 0.05% to 0.5%. Site dose
analyses are required to demonstrate that regulatory dose limits are
met using Technical Specification allowed penetration and system
bypass with an appropriate safety factor as an input to the
evaluation. Since the dose analyses have not been modified to credit
0.05% penetration and system bypass, this proposed change has no
effect on the dose analyses which demonstrate that the regulatory
limits are satisfied. Since the NRC regulatory limits must continue
to be met and the safety factor will not be changed by this proposed
Technical Specification change, this change does not involve a
significant reduction in the margin of safety.
Addition of Filter Test Face Velocities
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
This license amendment request proposes to add filter test face
velocity minimum values for the control room special ventilation
system, auxiliary building special ventilation system, spent fuel
pool special and inservice purge ventilation system and shield
building ventilation system. These ventilation systems are included
in the plant design to mitigate accident consequences and are not
assumed accident initiators, thus, this change does not involve a
significant increase in the probability of an accident. This change
will assure that the subject ventilation systems will perform within
their intended design ranges thus, this change assures that the
consequences of an accident are not increased.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
This proposed change does not alter the design, function, or
operation of any plant component and does not install any new or
different equipment. The malfunction of safety related equipment,
assumed to be operable in the accident analyses, would not be caused
as a result of the proposed Technical Specification change. No new
failure mode has been created and no new equipment performance
burdens are imposed. Therefore the possibility of a new or different
kind of accident from those previously analyzed has not been
created.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
This license amendment request proposes to add filter test face
velocity minimum values for the control room special ventilation
system, auxiliary building special ventilation system, spent fuel
pool special and inservice purge ventilation system and shield
building ventilation system. These additional Technical
Specification limits on system performance assures these ventilation
systems are tested and maintained within their designed function
limits and may increase the margin of safety for these systems.
Therefore this change does not involve a significant reduction in
the margin of safety.
Editorial and Administrative Changes
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
This license amendment request proposes editorial changes to
Technical Specification Section 5.5.9, including replacement of
ventilation system names with abbreviations and miscellaneous
changes associated with addition of a new paragraph to this section,
and proposes an administrative change to delete the Operating
License Additional Condition for each unit that relates to NRC
Generic Letter 99-02. Since these changes are editorial or
administrative, they do not change any plant operating limits or
technical requirements. Therefore these changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
This proposed change does not alter the design, function, or
operation of any plant component and does not install any new or
different equipment. The malfunction of safety related equipment,
assumed to be operable in the accident analyses, would not be caused
as a result of the proposed technical specification change. No new
failure mode has been created and no new equipment performance
burdens are imposed. Therefore, the possibility of a new or
different kind of accident from those previously analyzed has not
been created.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
This license amendment request proposes editorial changes to
Technical Specification Section 5.5.9, including replacement of
ventilation system names with abbreviations and miscellaneous
changes associated with addition of a new paragraph to this section,
and proposes an administrative change to delete the Operating
License Additional Condition for each unit that relates to NRC
Generic Letter 99-02. Since these changes are editorial or
administrative, they do not change any plant operating limits or
technical requirements. Therefore these 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 requests involve 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: L. Raghavan.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: March 11, 2003.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3.1.4, ``Rod Group Alignment
Limits,'' and TS 3.1.7, ``Rod Position Indication,'' to allow up to 1
hour of soak time following substantial rod movement during which
individual rod position indicators may not be within its limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
This license amendment request proposes to allow up to one hour
of soak time following substantial rod movement during which time
the rod position indication may be outside its limits. This would
allow an additional hour for rod position indication to be
inoperable or a control rod to be misaligned prior to entry into a
Technical Specification LCO [Limiting Condition for Operation]
Condition and Required Actions.
Rod position indication instrumentation is not an assumed
accident initiator and thus this change does not involve a
significant increase in the probability of an accident. Rod position
indication instrumentation provides information on control rod
position. Inoperable rod position indication instrumentation for an
additional hour does not make a rod misaligned. The consequences of
a rod misaligned for an additional hour are considered separately,
thus inoperable rod position indication instrumentation, by itself,
for an additional hour does not involve an increase in the
consequences of an accident.
This license amendment request may allow a misaligned rod to be
undetected for an additional hour. Plant safety analyses consider
two types of rod misalignment events, static misalignment and a
dropped rod. This license amendment request does not involve a
significant increase in the probability of a misaligned control rod
event because the one-hour time extension does not affect the
control rod drive system features, whose failure would result in
either type of misalignment. This proposed one-hour time extension
for a control rod to be misaligned does not involve a significant
increase in the
[[Page 18281]]
consequences of a rod misalignment event as follows. The analyses
show that a single dropped rod event, without any operator
intervention, does not result in any fuel pin failure, therefore the
rod drop event is not time dependent and an additional hour with the
misalignment undetected and unmitigated does not increase the
consequences of the event. Multiple rod drop events cause the
reactor to trip and therefore an additional hour would not have any
impact on this event.
In the static misalignment event, one or more control rods are
assumed to be statically misplaced from the allowed position. This
situation might occur if a rod were left behind when inserting or
withdrawing banks, or if a single rod were to be withdrawn. The
analysis of this event is bounded by modeling the most limiting
configuration which is the control banks at the full power insertion
limit except for a single control rod fully withdrawn. The analyses
show that, without any operator intervention, a single fully
withdrawn rod event does not result in any fuel pin failure,
therefore the static rod misalignment event is not time dependent
and an additional hour with the misalignment undetected and
unmitigated does not increase the consequences of the event.
Multiple rod misalignment events are bounded by the single rod
misalignment analyses and therefore an additional hour would not
have any impact on this event.
Therefore this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
This proposed change does not alter the design, function, or
operation of any plant component and does not install any new or
different equipment. The malfunction of safety related equipment,
assumed to be operable in the accident analyses, would not be caused
as a result of the proposed technical specification change. No new
failure mode has been created and no new equipment performance
burdens are imposed. Therefore the possibility of a new or different
kind of accident from those previously analyzed has not been
created.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
This license amendment request proposes to allow up to one hour
of soak time following substantial rod movement during which time
the rod position indication may be outside its limits. This would
allow an additional hour for rod position indication instrumentation
to be inoperable or a control rod to be misaligned prior to entry
into a Technical Specification LCO Condition and Required Actions.
The rod position indication system is an instrumentation system
that provides indication to the operators that a control rod may be
misaligned. Inoperable individual rod position indication
instrumentation does not by itself in any way harm or impact reactor
operation. Inoperable rod position indication instrumentation may
impair the ability of the operators to detect a misaligned rod. The
impact of inoperable rod position indication instrumentation may be
offset by availability of other indications that a rod is misaligned
such as nuclear instrumentation indication that reactor power has
shifted to one side of the core or thermocouple indication that the
core temperatures increased in one region of the core and/or
decreased in another region of the core.
The Prairie Island staff is not aware of a misaligned control
rod in more than 50 reactor-years of plant operation. The likelihood
of a misaligned rod at Prairie Island is small and the likelihood of
a misaligned rod coincident with inoperable rod position indication
during the allowed one-hour extension is smaller.
The addition of one hour soak time for the rod position
indication instrumentation will allow the operators and engineers to
focus on monitoring the reactor performance without unnecessary
entry into LCO Conditions and Required Actions with the concomitant
administrative activities. Thus, these changes may enhance plant
safety and reliability of equipment.
In conclusion, the proposed addition of an LCO Note in LCO 3.1.4
and 3.1.7 does not involve a significant reduction in the margin of
safety because rod position indication instrumentation inoperability
by itself does not impact plant safety, rod misalignment is
unlikely, there may be other indications of rod misalignment, rod
misalignment coincident with rod position indication instrumentation
inoperability within the one hour extension is unlikely, and plant
safety may be enhanced by avoiding unnecessary LCO Condition entry.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: March 19, 2003.
Description of amendment request: The proposed amendments would
revise the Technical Specification (TS) 5.3, ``Plant Staff
Qualifications.'' The proposed amendments would revise requirements
that have been superseded based on licensed operator training programs
being accredited by the National Academy for Nuclear Training (NANT)
and promulgation of the revised 10 CFR part 55, ``Operators'
Licenses,'' which became effective on May 26, 1987.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed Technical Specification change is an administrative
change to clarify the current requirements for licensed operator
qualifications and the licensed operator training program. With this
change, the Technical Specifications continue to meet the current
requirements of 10 CFR [Part] 55.
Although licensed operator qualifications and training may have
an indirect impact on accidents previously evaluated, the NRC
considered this impact during the rulemaking process, and by
promulgation of the revised 10 CFR [Part] 55 rule, concluded that
this impact remains acceptable as long as the licensed operator
training programs are certified to be accredited and are based on a
systems approach to training. The Prairie Island Nuclear Generating
Plant licensed operator training program is accredited by the
National Academy for Nuclear Training and is based on a systems
approach to training. The proposed Technical Specification change
takes credit for the National Academy for Nuclear Training
accreditation of the licensed operator training program. The
Technical Specification requirements for all other plant staff
qualifications remain unchanged.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed Technical Specification change is an administrative
change to clarify the current requirements for licensed operator
qualifications and the licensed operator training program and to
conform to the revised 10 CFR [Part] 55.
As discussed above, although licensed operator qualifications
and training may have an indirect impact on the possibility of a new
or different kind of accident from any accident previously
evaluated, the NRC considered this impact during the rulemaking
process, and by promulgation of the revised rule, concluded that
this impact remains acceptable as long as licensed operator training
programs are certified to be accredited and based on a systems
approach to training. As previously noted, the Prairie Island
Nuclear Generating Plant licensed operator training program is
accredited by the National Academy for Nuclear Training and is based
on a systems approach to training. The proposed Technical
[[Page 18282]]
Specification change takes credit for the National Academy for
Nuclear Training accreditation of the licensed operator training
program. The Technical Specification requirements for all other
plant staff qualifications remain unchanged.
Additionally, the proposed Technical Specification change does
not affect plant design, hardware, system operation, or procedures.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed Technical Specification change is an administrative
change to clarify the current requirements applicable to licensed
operator qualifications and the licensed operator training program.
With this change the Technical Specifications continue to be
consistent with the requirements of 10 CFR [Part] 55. The Technical
Specification qualification requirements for all other plant staff
remain unchanged.
Licensed operator qualifications and training can have an
indirect impact on a margin of safety. However, the NRC considered
this impact during the rulemaking process, and by promulgation of
the revised 10 CFR [Part] 55, determined that this impact remains
acceptable when licensees maintain a licensed operator training
program that is accredited and based on a systems approach to
training. As noted previously, the Prairie Island Nuclear Generating
Plant licensed operator training program is accredited by the
National Academy for Nuclear Training and is based on a systems
approach to training.
The NRC has concluded, as stated in NUREG-1262, ``Answers to
Questions at Public Meetings Regarding Implementation of Title 10,
Code of Federal Regulations, Part 55 on Operators' Licenses,'' that
the standards and guidelines applied by the Institute for Nuclear
Power Operations' National Academy for Nuclear Training in their
training accreditation program are equivalent to those put forth or
endorsed by the NRC. As a result, maintaining a National Academy for
Nuclear Training accredited, systems approach based licensed
operator training program is equivalent to maintaining an NRC
approved licensed operator training program which conforms with
applicable NRC Regulatory Guides or NRC endorsed industry standards.
The margin of safety is maintained by virtue of maintaining the
National Academy for Nuclear Training accredited licensed operator
training program.
In addition, the NRC published NRC Regulatory Issue Summary
2001-01, ``Eligibility of Operator License Applicants,'' dated
January 18, 2001, ``to familiarize addressees with the NRC's current
guidelines for the qualification and training of reactor operator
(RO) and senior operator (SO) license applicants.'' This document
again acknowledges that the Institute for Nuclear Power Operations'
National Academy for Nuclear Training guidelines for education and
experience, outline acceptable methods for implementing the NRC's
regulations in this area.
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 requests involve 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: L. Raghavan.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: February 6, 2003.
Description of amendment requests: The proposed license amendments
would revise Surveillance Requirements (SRs) 3.3.1.2 and 3.3.1.3 of TS
3.3.1, ``Reactor Trip System Instrumentation,'' of the Diablo Canyon
Technical Specifications. The change to SR 3.3.1.2 is consistent with
NRC-approved Industry/Technical Specifications Task Force (TSTF)
Standard Technical Specification Change Traveler, TSTF-371. The change
to SR 3.3.1.3 is editorial in nature.
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 change to Technical Specifications (TS)
Surveillance Requirement (SR) 3.3.1.2 and SR 3.3.1.3 is consistent
with the NRC approved Industry/Technical Specifications Task Force
Standard Technical Specification Change Traveler, TSTF-371, and
NUREG-1431, ``Standard Technical Specifications, Westinghouse
Plants,'' Revision 2.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since there are
no hardware changes. The reactor trip system (RTS) instrumentation
will be unaffected. Protection systems will continue to function in
a manner consistent with the plant design basis. All design,
material, and construction standards that were applicable prior to
the request are maintained.
The probability and consequences of accidents previously
evaluated in the Updated Final Safety Analysis Report (UFSAR) are
not adversely affected because the change to the nuclear
instrumentation system (NIS) power range channel daily surveillance
assures the conservative response of the channel even at part-power
levels.
The proposed change modifies the NIS power range channel daily
surveillance requirement to help assure the NIS power range
functions are tested in a manner consistent with the safety analysis
and licensing basis.
The proposed change will not affect the probability of any event
initiators. There will be no degradation in the performance of, or
an increase in the number of challenges imposed on, safety-related
equipment assumed to function during an accident situation. There
will be no change to normal plant operating parameters or accident
mitigation performance.
The proposed change will not alter any assumptions or change any
mitigation actions in the radiological consequence evaluations in
the USAR.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There is no hardware change or change in the method by which any
safety-related plant system performs its safety function. This
change will not affect the normal method of plant operation or
change any operating parameters. No performance requirements or
response time limits will be affected. The NIS power range high trip
setpoint adjustment requirements, prior to adjusting indicated power
in a decreasing power direction, will ensure the reactor power level
is consistent with assumptions made in the safety analysis and
licensing basis.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed change. There will be no adverse effect or
challenges imposed on any safety-related system as a result of the
change.
This amendment does not alter the design or performance of the
Eagle 21 System, NIS, or Solid State Protection System used in the
plant protection systems.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change requires a revision to the criteria for
implementation of NIS power range channel adjustments based on
secondary power calorimetric calculations; however, the change does
not eliminate any RTS surveillances or alter the frequency of
surveillances required by the Technical Specifications. The revision
to the criteria for implementation of the daily surveillance will
have a conservative effect on the performance of the NIS power range
channels, particularly at part-power conditions. The nominal trip
setpoints specified in the Technical Specification Bases and the
safety analysis
[[Page 18283]]
limits assumed in the transient and accident analyses are unchanged.
None of the acceptance criteria for any accident analysis is
changed.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions. There will be no impact on the overpower
limit, departure from nucleate boiling ratio limits, heat flux hot
channel factor (FQ), nuclear enthalpy rise hot channel factor (FDH),
loss of coolant accident peak cladding temperature, peak local power
density, or any other margin of safety. The radiological dose
consequence acceptance criteria listed in the Standard Review Plan
will continue to be met.
The imposition of appropriate surveillance testing requirements
will not reduce any margin of safety since the change will assure
that safety analysis assumptions on reactor power are verified on a
periodic frequency.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: February 28, 2003.
Description of amendment requests: The proposed license amendments
would revise Technical Specification (TS) 3.3.1, ``Reactor Trip System
(RTS) Instrumentation,'' to add Surveillance Requirement (SR) 3.3.1.16
to function 3.a, ``Power Range Neutron Flux Rate-High Positive Rate
Trip,'' in Table 3.3.1-1. The amendments would also eliminate periodic
pressure sensor response time testing (RTT) and periodic protection
channel RTT.
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.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since there are
no hardware changes.
The design of the Reactor Trip System (RTS) instrumentation,
specifically the positive flux rate trip (PFRT) function, will be
unaffected. The reactor protection system will continue to function
in a manner consistent with the plant design basis. All design,
material, and construction standards that were applicable prior to
the request are maintained.
The proposed change imposes additional surveillance requirements
to assure safety-related structures, systems, and components are
verified to be consistent with the safety analysis and licensing
basis. In this specific case, a response time verification
requirement will be added to the PFRT function.
The Technical Specification Bases changes do not result in a
condition where the design, material, or construction standards that
were applicable prior to change are altered. The same RTS and
engineered safety features actuation system instrumentation is being
used; the time response allocations/modeling assumptions in the
Updated Final Safety Analysis Report (UFSAR) Chapter 15 analyses are
still the same; only the method of verifying time response is
changed. The proposed change will not change any system interface
and could not increase the likelihood of an accident since these
events are independent of this change.
The proposed change will not affect the probability of any event
initiators. There will be no degradation in the performance of, or
an increase in the number of challenges imposed on safety-related
equipment assumed to function during an accident situation. There
will be no change to normal plant operating parameters or accident
mitigation performance.
The proposed activity will not change, degrade or prevent
actions or alter any assumptions previously made in evaluating the
radiological consequences of an accident described in the UFSAR.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There are no hardware changes nor are there any changes in the
method by which any safety-related plant system performs its safety
function. This change will not affect the normal method of plant
operation or change any operating parameters. No performance
requirements will be affected; however, the proposed change does
impose additional surveillance requirements for the PFRT function.
These additional requirements are consistent with assumptions made
in the safety analysis and licensing basis.
This change does not alter the performance of the process
protection racks, nuclear instrumentation, and logic systems used in
the plant protection systems. These systems will still have their
response time verified by test before being placed in operational
service. Changing the method of verifying instrument response for
these systems (assuring equipment operability) from time response
testing to channel and calibration checks will not create any new
[accident] initiators or scenarios. Periodic surveillance of these
systems will continue and may be used to detect degradation that
could cause the response time characteristic to exceed the total
allowance. The total response time allowance for each function
bounds all degradation that cannot be detected by periodic
surveillance.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of this change. There will be no adverse effects or challenges
imposed on any safety-related system as a result of this change.
Therefore the proposed change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions. There will be no impact on the overpower
limit, departure from nucleate boiling ratio limits, heat flux hot
channel factor, nuclear enthalpy rise hot channel factor, loss of
coolant accident peak cladding temperature, peak local power
density, or any other margin of safety. The radiological dose
consequence acceptance criteria listed in the Standard Review Plan
will continue to be met.
The safety analysis limits assumed in the transient and accident
analyses are unchanged. None of the acceptance criteria for any
accident analysis are changed. The imposition of additional
surveillance requirements maintains the margin of safety by assuring
that the affected safety analysis assumptions on equipment response
time are verified on a periodic frequency.
This change does not affect the total system response time
assumed in the safety analysis. The periodic system response time
verification method for the process protection racks, nuclear
instrumentation, and logic systems are modified to allow use of
engineering data. The method of verification still provides
assurance that the total system response is within that defined in
the safety analysis, since calibration tests will continue to be
performed and may be used to detect any degradation which might
cause the response time to exceed the total allowance. The total
response time allowance for each function bounds all degradation
that cannot be detected by periodic surveillance. Based on the
above, it is concluded that the proposed change does not result in a
significant reduction in margin with respect to plant safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c)
are satisfied. Therefore, the NRC staff proposes to determine that
the amendment requests involve no significant hazards consideration.
[[Page 18284]]
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas
and Electric Company, P.O. Box 7442, San Francisco, California
94120.
NRC Section Chief: Stephen Dembek.
Pacific Gas and Electric Company, Docket No. 50-323, Diablo Canyon
Nuclear Power Plant, Unit No. 2, San Luis Obispo County, California
Date of amendment request: March 3, 2003.
Description of amendment request: The proposed license amendment
would revise Technical Specification (TS) 5.5.9, ``Steam Generator
Tube Surveillance Program,'' and TS 5.6.10, ``Steam Generator Tube
Inspection Report,'' for Diablo Canyon Power Plant (DCPP) Unit 2, to
apply a probability of detection (POD) of 1.0 to the bobbin
indication in the steam generator (SG) 4 tube at row 44, column 45
at the second tube support plate (TSP) on the hot leg side (R44C45-
2H) for the beginning of cycle (BOC) voltage distribution for the
DCPP Unit 2 BOC Cycle 12 operational assessment.
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 use of probability of detection (POD) of 1.0 for the bobbin
indication in the Diablo Canyon Power Plant (DCPP) Unit 2 steam
generator (SG) 4 tube at row 44, column 45 at the second tube
support plate (TSP) on the hot leg side (R44C45-2H) for the
beginning of cycle (BOC) voltage distribution for the DCPP Unit 2
BOC cycle 12 operational assessment does not increase the
probability of an accident. Based on industry and plant specific
bobbin detection data for outside diameter stress corrosion cracks
(ODSCC) within the SG tube support plate region, large voltage
bobbin indications, such as those the size of indication R44C45-2H,
can be detected with 100 percent certainty. Since large voltage
ODSCC bobbin indications within the SG TSP can be detected, they
will not be left in service, and therefore these indications should
not be included in the voltage distribution for the purpose of
operational assessments. Therefore, these large voltage indications
will not result in an increase in the probability of a steam
generator tube rupture (SGTR) accident or an increase in the
consequences of a SGTR or main steam line break (MSLB) accident.
Therefore, the proposed changes will not result in 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.
The use of a POD of 1.0 for the DCPP Unit 2 R44C45-2H bobbin
indication for the BOC voltage distribution for the DCPP Unit 2 BOC
cycle 12 operational assessment concerns the SG tubes and can only
affect the SGTR accident. Since the SGTR accident is already
considered in the Final Safety Analysis Report Update, there in [is]
no possibility to create a design basis accident which has not been
previously evaluated.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The use of POD of 1.0 for the DCPP Unit 2 R44C45-2H bobbin
indication for the BOC voltage distribution for the DCPP Unit 2 BOC
cycle 12 operational assessment does not involve a significant
reduction in a margin of safety. The applicable margin of safety
potentially impacted is the Technical Specification 5.6.10, ``Steam
Generator Tube Inspection Report,'' projected end-of-cycle leakage
for a MSLB accident and the projected end-of-cycle probability of
burst. Based on industry and plant specific bobbin detection data
for ODSCC within the SG tube support plate region, large voltage
bobbin indications, such as those the size of indication R44C45-2H,
can be detected with 100 percent certainty and will not be left in
service. Therefore these indications should not be included in the
voltage distribution for the purpose of operational assessments.
Therefore, these large voltage indications will not result in a
significant increase in the actual end-of-cycle leakage for a MSLB
accident or the actual end-of-cycle probability of burst.
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 requests involve no significant hazards consideration.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas
and Electric Company, P.O. Box 7442, San Francisco, California
94120.
NRC Section Chief: Stephen Dembek.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: February 14, 2003.
Description of amendment request: The proposed amendment would
extend the surveillance test intervals and allowed out-of-service
times for the end-of-cycle recirculation pump trip instrumentation.
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?
Response: No.
The proposed amendment would extend the allowed out-of-service
times (AOTs) and surveillance test intervals (STIs) for the end of
cycle recirculation pump trip (EOC-RPT) instrumentation system. No
changes are being made to any EOC-RPT instrumentation setpoints or
components. The effect of the proposed changes is to reduce the
potential for unnecessary plant scrams or transients. The proposed
changes were evaluated in General Electric Company Topical Report
GENE-770-06-1-A which concluded that they do not result in a
degradation in overall plant safety.
Since the proposed changes do not affect any accident initiator,
and since the EOC-RPT instrumentation will remain capable of
performing its design function, 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.
Extending the AOTs and STIs for the EOC-RPT instrumentation does
not change the design function or operation of any plant equipment.
Additionally, no new modes of plant operation are involved with
these changes.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
No changes are being made to any plant instrumentation setpoints
or to the required level of redundancy. The proposed changes were
evaluated in General Electric Company Topical Report GENE-770-06-1-
A, which concluded that they do not result in a degradation in
overall plant safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: January 29, 2003.
Description of amendment request: The licensee proposed
administrative and editorial changes to the Salem Nuclear Generating
Station (Salem), Unit No. 1 and Unit No. 2 Technical Specifications
(TSs) as follows: (1) The second equation in Salem Unit No. 2 TS
[[Page 18285]]
Limiting Condition for Operation 3.2.2 on page 3/4 2-5 will be revised;
(2) Salem Unit No. 2 TS Table 3.3-6 will be revised to indicate that
one operable channel of containment air particulate activity reactor
coolant system (RCS) leakage detection instrumentation is required for
operation in Modes 1 through 4; (3) Salem Unit No. 1 TS 3/4.7.6 Action
Statements ``d.'' (for Modes 1, 2, 3 and 4) and ``e.'' (for Modes 5 and
6) will be revised to refer to Action 25 in TS Table 3.3-6; and (4)
Salem Unit No. 2 TS 3/4.7.6 Action Statements ``d.'' (for Modes 1, 2, 3
and 4) and ``e.'' (for Modes 5 and 6) will be revised to refer to
Action 28 in TS Table 3.3-6.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the TSs are administrative or editorial
in nature and do not change the intent of any Technical
Specification requirement. No changes are being made to any plant
systems, structures or components (SSCs).
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 administrative and editorial changes to the TSs do
not change the design function or operation of any plant equipment.
Additionally, no new modes of plant operation are involved with
these changes.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes are administrative and editorial
corrections to the TSs that do not affect the ability of plant SSCs
to perform their design basis accident functions. 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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
Southern California Edison Company, et al., Docket No. 50-206, San
Onofre Nuclear Generating Station, Unit 1, San Diego County, California
Date of amendment request: March 11, 2003.
Description of amendment request: The amendment application
requests a revision to the Unit 1 defueled Technical Specifications
administrative controls section to propose changes in organizational
responsibilities. Specifically, the proposed change identifies that the
Vice President, Engineering & Technical Services would be responsible
for decommissioning activities. Additionally, the Station Manager would
be designated as having approval authority for activities within the
Station Manager's organization.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
No. This is a request to revise the San Onofre Nuclear
Generating Station, Unit 1 permanently defueled technical
specifications administrative controls. The proposed administrative
changes are due to a realignment of the Unit 1 Decommissioning
Project into the Engineering & Technical Services organization and
the establishment of the Station Manager position within the Nuclear
Generation organization. The proposed changes identify the Vice
President, Engineering & Technical Services to be responsible for
decommissioning activities and provides the Station Manager the
opportunity to approve procedures and changes to procedures and
changes to the Process Control Program that are under the Station
Manger's responsibility. Therefore, the proposed changes do not
involve an increase in the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different type of accident
from any accident previously evaluated?
No. The proposed changes are administrative. Therefore, the
proposed changes do not involve the possibility of a new or
different type of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety?
No. The proposed changes are administrative. Therefore, the
proposed changes do not involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis. These
administrative changes do not affect the design or operation of the
facility 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: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Acting Section Chief: Mark Thaggard.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendments request: March 25, 2003.
Description of amendments request: The proposed amendments would
revise Technical Specification 3.5.2, ``ECCS--Operating,'' Surveillance
Requirement (SR) 3.5.2.5. Specifically, the proposed change would
replace the requirement to verify specific surveillance test values for
the Emergency Core Cooling System (ECCS) pumps with the requirement to
verify the developed head for each ECCS pump in accordance with the
Inservice Testing Program. This new requirement is identical to SR
3.5.2.4 in NUREG-1432, ``Standard Technical Specifications, Combustion
Engineering Plants,'' Revision 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Deleting the specific surveillance test values for Emergency
Core Cooling System (ECCS) pumps from Surveillance Requirement (SR)
3.5.2.5 does not affect the probability of occurrence or
consequences of an accident previously evaluated because ECCS pumps
are for accident mitigation and do not contribute to initiation of
accidents. Periodic surveillance testing of the ECCS pumps in
accordance with the Inservice Testing (IST) program provides
assurance that the pumps will perform as assumed in the safety
analysis. There is no change to the safety analysis.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[[Page 18286]]
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
ECCS pumps are for accident mitigation and do not contribute to
accident initiation. The ECCS system will still be verified capable
of meeting its emergency core cooling and IST requirements. There is
no change to the safety analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
There is no change to the safety analysis. Testing of the ECCS
pumps as required by the IST Program combined with the existing
Technical Specification 3.5.2--``ECCS--Operating'' surveillance
requirements ensure that the ECCS requirements remain met without a
significant reduction in a margin of safety. Therefore, there is no
significant reduction in a margin of safety.
Based on the above, SCE [Southern California Edison Company]
concludes that the proposed amendments present 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.
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
amendments request involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.
Tennessee Valley Authority, Docket Nos. 50-327 and 328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: March 13, 2003.
Description of amendment request: The proposed amendments would
modify the Sequoyah Nuclear Plant, Units 1 and 2, Operating Licenses
DPR-77 and DPR-79. This proposed request provides Technical
Specification (TS) change 03-01 that would revise the limiting
condition for operation for TS Section 3.5.1, ``Cold Leg Injection
Accumulators'' and TS Section 3.5.5, ``Refueling Water Storage Tank.''
This revision would modify the single boron concentration requirement
by inserting a table that defines the minimum and maximum amount of
boron that is required for accident mitigation based on the number of
tritium producing rods in the core.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change modifies the required boron concentration
for the cold leg accumulators (CLAs) and refueling water storage
tank (RWST). The proposed values have been verified to maintain the
required accident mitigation safety function for the CLAs and RWST.
The CLAs and RWST safety function is to mitigate accidents that
require the injection of borated water to cool the core and to
control reactivity. These functions are not potential sources for
accident generation and the modification of the boron concentration
that supports event mitigation will not increase the potential for
an accident. Therefore, the possibility of an accident is not
increased by the proposed changes. The boron levels for this change
are based on the number or tritium producing rods in the core. As
the number of rods is increased the need for additional shutdown
boron also increases. This effect has been evaluated with the same
methodology utilized for previous NRC approved amendments associated
with tritium production. This methodology ensures that the impact of
tritium producing rods is adequately compensated for by the required
boron concentrations and has been incorporated into the proposed
revision. Since the boron levels will continue to maintain the
safety function of the CLAs and RWST in the same manner as currently
approved, the consequences of an accident is not increased by the
proposed changes.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change only modifies boron concentrations for
accident mitigation functions of the CLAs and RWST. These functions
do not have a potential to generate accidents as they only serve to
perform mitigation functions associated with an accident. The
proposed requirements will maintain the mitigation function in an
identical manner as currently approved. There are no plant equipment
or operational changes associated with the proposed revision other
than the adjustment of the boron level in the CLAs and RWST.
Therefore, since the CLA and RWST functions are not altered and the
plant will continue to operate without change, the possibility of a
new or different kind of an accident is not created.
3. Does the proposed change involve a significant reduction in a
margin of safety?
This change proposes boron concentration requirements that
support the accident mitigation functions of the CLAs and RWST
equivalent to the currently approved limits. The proposed change
does not alter any plant equipment or components and does not alter
any setpoints utilized for the actuation of accident mitigation
system or control functions. The proposed boron values have been
verified to provide the same level of reactivity control for
accident mitigation. Therefore, the proposed change will 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 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority (TVA), Docket No. 50-390, Watts Bar Nuclear
Plant, Unit 1, Rhea County, Tennessee
Date of amendment request: March 12, 2003.
Description of amendment request: The proposed amendment would
revise the Updated Final Safety Analysis Report (UFSAR) and the
Technical Specification Bases. The revision would update the quality
assurance criteria and the basis for the seismic qualification of the
ducting installed as part of the suspended ceiling air delivery system
in the main control room (MCR).
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. The design function of the MCR ducting system is to support
pressurization and cooling of the control room during normal and
accident conditions. The MCR ducting is a passive plant feature and
does not act as an accident initiator. Consequently, the changes in
the MCR ducting system and suspended ceiling quality assurance (QA)
requirements and qualification methodology do not result in an
increase in the probability of an accident previously evaluated.
For the principal design basis accidents, Loss of Coolant
Accident (LOCA), Internal Flood, Steam Generator Tube Rupture
(STGR), Main Steam Line Break (MSLB), etc., the integrity of the MCR
HVAC [heating, ventilation, and air conditioning] system, including
the suspended ceiling, will not be compromised. These accidents do
not have a structural effect on the MCR. This means that for
postulated radiological or toxic chemical accidents, the ability to
both pressurize and
[[Page 18287]]
maintain MCR temperatures within the design limits is unaffected by
the limited QA and newly defined seismic requirements for the air
delivery components.
An accident that involves a fire that affects the MCR or the
habitability of the MCR was not a consideration for the
qualification of the air distribution components. A fire of this
nature will result in plant operation from the Auxiliary Control
Room which is supported by a separate heating, ventilation and air
conditioning (HVAC) supply system.
An earthquake (including the Design Basis SSE [safe shutdown
earthquake]) is the only event for which the design basis for the
MCR HVAC and suspended ceiling is potentially challenged. A seismic
qualification report by an industry seismic expert concludes that
the air delivery components will remain in place, will retain their
structural integrity such that flow will not be impeded, and the
pressure boundary will not be lost during and subsequent to a design
basis seismic event. Further, as assured by TVA's qualification
report, the suspended ceiling will remain in place during and
subsequent to a seismic event or accident. Thus, the revised QA and
seismic qualification requirements for the MCR air delivery
components and suspended ceiling will not result in loss of safety
function for any design basis accident or event. Consequently, the
accident dose consequences as previously evaluated in the UFSAR are
not affected by the proposed license amendment.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The MCR air delivery components addressed by the proposed
amendment are not an accident initiator and therefore, failure of
these components will not initiate a design basis accident. In
addition, the subject air delivery components and suspended ceiling
have been seismically qualified, as previously discussed, and a
determination has been made that they will not fail during a design
basis accident. Therefore, the air delivery components and suspended
ceiling will continue to perform their safety function during normal
and accident conditions. Consequently, this activity does not create
a possibility of a new or different type of accident than any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The changes addressed in TVA's proposed amendment are
associated with changes in QA requirements and seismic qualification
methodology for safety related air delivery components and for the
suspended ceiling. The change does not affect specific HVAC
equipment safety limits, design limits, set points, or other
critical parameters. In addition, the new seismic analysis
methodology and limited QA requirements ensure that these components
will continue to perform their safety functions during normal and
accident conditions. The previously implied margin of safety against
structural or functional failure of the air delivery components or
suspended ceiling during and after a design basis SSE has not been
reduced. Consequently, the MCR HVAC system or suspended ceiling
margin of safety has not been significantly reduced by this proposed
amendment.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: March 24, 2003.
Description of amendment request: The proposed amendment would
revise the design and licensing basis failure modes and effects
analysis for specific valves in the essential raw cooling water system,
component cooling water system, and control air system. Tennessee
Valley Authority has identified a condition where containment
integrity, accident flood levels, and sump boron concentrations
subsequent to a high-energy line break events could not be assured
automatically as stated in the updated final safety analysis report
(UFSAR). In certain postulated events, manual actions may be required
using equipment not currently evaluated in the UFSAR.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously
evaluated[?]
Response: No.
The manual actions required by this change are only needed after
a high energy line break (HELB) accident, such as a loss-of-coolant-
accident (LOCA), main steam line break (MSLB), feedwater line break
accidents, etc., has occurred inside containment and a single
failure of an outboard containment isolation valve to close has
occurred on one of four specific lines inside containment. In this
event, the manual actions ensure containment is isolated, which is
consistent with the current design. Consequently, the manual actions
of isolating the air and water lines after an accident do not affect
the frequency of any accident previously evaluated in the Updated
Final Safety Analysis Report (UFSAR).
The UFSAR currently indicates that the containment vessel design
and the containment isolation system automatically ensure
containment integrity is maintained and thus ensure that release of
radioactive material from containment remains below allowable limits
during and subsequent to an accident. Current UFSAR Failure Modes
and Effects Analysis (FMEA) for the affected essential raw cooling
water (ERCW), component cooling system (CCS), and control air system
(CAS) valves indicate a single failure of the outboard containment
isolation valve in conjunction with a concurrent accident and
consequential (due to interaction) failure of the system piping
inside containment, has no adverse effect on the plant; thus,
containment integrity is ensured automatically. This change revises
these evaluations to indicate manual actions are required to ensure
containment integrity in the event of an HELB and single failure of
an outboard containment isolation valve. Evaluations have been
performed to ensure adequate instrumentation and time is available
to recognize the need and to manually isolate an affected line
subsequent to an HELB if the outboard containment isolation valve
does not close. The emergency procedures have been revised that
require manual actions to be performed to isolate CAS, ERCW, and CCS
and to open and close a post accident sampling facility (PASF)
cooling water supply valve. The Operations Staff has confirmed that
the subject containment lines can be isolated within the allowable
time and without exceeding the dose limitations as required by 10
CFR [Part] 50, Appendix A, General Design Criteria (GDC) 19,
``Control Room.''
Evaluations have indicated that adequate instrumentation, time,
and staffing are available to manually isolate the lines into
containment. Operator actions are achievable and can be accomplished
without heroic actions. Therefore, containment integrity from
overpressurization or flooding is maintained within the current
design basis analysis, and the radiological consequences of an
accident will not be increased by this change. Consequently, 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.
This change implements manual actions to isolate four specific
containment lines in lieu of automatic containment isolation for
previously identified accidents. The manual actions are required to
maintain containment integrity from overpressurization, containment
flood levels, sump pH levels, and emergency core cooling system
(ECCS) water boron concentrations subsequent to an HELB inside
containment concurrent with a single failure of an outboard
containment isolation valve on a CAS, ERCW, or CCS line. The UFSAR
FMEA evaluations will be revised by this proposed change to include
the failure modes and associated manual actions.
NRC Information Notice (IN) 97-78, ``Crediting of Operator
Actions in Place of Automatic Actions and Modifications of
[[Page 18288]]
Operator Actions, including Response Times,'' provided guidance to
the industry concerning use of operator actions in place of
automated system or component actuation. IN 97-78 states: In those
instances where licensees consider temporary or permanent changes to
the facility which credit operator actions, the NRC has relied on
the guidance provided in * * * ANSI/ANS 58.8, ``Time Response Design
Criteria for Safety-related Operator Actions,'' * * * for evaluating
such changes. The American Nuclear Society (ANS)-58.8, establishes
the requirements for safety-related operator actions, which are
summarized as follows: (1) The specific operator actions required,
(2) the potentially harsh or inhospitable environmental conditions
expected, (3) ingress/egress paths taken by the operators to
accomplish functions, (4) procedural guidance for required actions,
(5) operator training and qualifications to carry out actions, (6)
any additional support personnel and/or equipment to carry out
actions, (7) information required by the control room staff to
determine whether action is required, including qualified
instrumentation to diagnose the situation and to verify that the
action is successfully, (8) ability to recover from credible errors
in performance of manual actions, and the expected time required to
make such a recovery, and (9) consideration of risk significance of
operator actions.
The manual actions implemented by this change can be completed
within the guidance and criteria provided in IN 97-78 and ANS-58.8.
Consequently, the manual actions can be credited in the mitigation
of the specific accidents. With credit for the manual actions to
isolate the affected lines subsequent to an accident inside
containment, the type of accidents and consequences currently
evaluated in the UFSAR, remains the same. Therefore, the proposed
change does not create the possibility of new or different kinds of
accidents from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety[?]
Response: No.
This change establishes requirements for manual actions to
isolate one air line and three water lines subsequent to an accident
inside containment concurrent with a single failure of a containment
isolation valve to close. The manual actions ensure air or water
cannot continue to enter containment with a single failure of an
outboard containment isolation valve when the line pressure boundary
inside containment is lost due to an accident and associated pipe
interactions. The safety-related configuration of the lines
(outboard motor operated valve and inboard check valve) continues to
ensure the containment environment is automatically prevented from
exiting the line to outside the containment. Safety-related
instrumentation is available to inform operators that the manual
actions are required, and operators have been trained in the
requirements for addressing the failures of valves to close. In
addition, adequate time and resources are available to perform the
manual actions. The manual actions meet the criteria for safety-
related operator actions contained in NRC IN 97-78 and ANS-58.8.
Further, the proposed change to allow credit for the manual actions
does not affect the offsite and Main Control Room dose consequences
of the accidents currently reported in UFSAR Chapter 15, Accident
Analyses. 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: Allen G. Howe.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: March 6, 2003.
Brief description of amendments: Technical Specifications Section
1.1 ``Definitions'' for Engineered Safety Feature (ESF) Response Time
and Reactor Trip System (RTS) Response Time require U.S. Nuclear
Regulatory Commission (NRC) review and approval of any methodology used
to allocate response times in lieu of measuring them. The application
requests NRC review and approval of a topical report to allow the use
of allocated signal processing and actuation logic response times in
the overall verification of the protection system channel response
time.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not result in a condition where the
design, material, and construction standards that were applicable
prior to the change are altered. The same RTS and ESFAS [Engineered
Safety Feature Actuation System] instrumentation are being used and
the time response allocations and modeling assumptions in the
Chapter 15 safety analysis are unchanged. Only the method of
verifying the time response is changed. The proposed change will not
modify any system interface and could not increase the likelihood of
an accident since these events are independent of this change. The
proposed activity will not change, degrade, or prevent actions or
alter any assumptions previously made in evaluating the radiological
consequences of an accident described in the FSAR [Final Safety
Analysis Report]. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not alter the performance of the
process protection racks, the nuclear instrumentation, or the logic
systems used in the plant protection systems. Periodic surveillance
of these systems will continue and may be used to detect degradation
that could cause the response time characteristics to exceed the
total allowance. Changing the method of periodically verifying
instrument response for these systems from response time testing to
calibration and channel checks will not create any new accident
initiators or scenarios. Periodic surveillance of these systems will
continue and may be used to detect degradation that could cause the
response time characteristic to exceed the total allowance. The
total time response allowance for each function bounds all
degradation that cannot be detected by periodic surveillance.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not affect the total system response
time assumed in the safety analysis. The periodic response time
verification method for the Process protection racks, the nuclear
instrumentation and the logic systems is modified to allow the use
of actual test data or engineering data. The method of verification
still provides assurance that the total system response time is
within that defined in the safety analysis, since calibration tests
will continue to be performed and may be used to detect any
degradation which might cause the response time to exceed the total
allowance. The total response time allowance for each function
bounds all degradation that cannot be detected by
[[Page 18289]]
periodic surveillance. Therefore the proposed change does 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: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: March 18, 2003.
Brief description of amendments: The proposed amendment would
delete certain of the Surveillance Requirements in Technical
Specification 3.6.3 entitled ``Containment Isolation Valves.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Do the proposed changes involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since there are
no hardware changes. Protection systems will continue to function in
a manner consistent with the plant design basis. All design,
material, and construction standards that were applicable prior to
the request are maintained.
The probability and consequences of accidents previously
evaluated in the FSAR [Final Safety Analysis Report] are not
adversely affected.
The proposed changes will not involve a significant increase in
the probability of any event initiators. There will be no
degradation in the performance of, or an increase in the number of
challenges imposed on, safety-related equipment assumed to function
during an accident situation. There will be no change to normal
plant operating parameters or accident mitigation performance.
The proposed changes will not alter any assumptions or change
any mitigation actions in the radiological consequence evaluations
in the FSAR.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(2) Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve any physical alteration of
the units. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. There
are no setpoints at which protective or mitigative actions are
initiated that are affected by the proposed change. The proposed
change will not alter the manner in which equipment operation is
initiated, nor will the function demands on credited equipment be
changed. No alteration in the procedures, which ensure the unit
remains within analyzed limits, is proposed, and no change is being
made to procedures relied upon to respond to an off-normal event. As
such, no new failure modes are being introduced. The proposed change
does not alter assumptions made in the safety analyses.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
(3) Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will not adversely affect operation of plant
equipment and will not result in a change to the setpoints at which
protective actions are initiated. None of the acceptance criteria
for any accident analysis is changed. There will be no effect on the
manner in which safety limits or limiting safety system settings are
determined nor will there be any effect on those plant systems
necessary to assure the accomplishment of protection functions.
There will be no impact on the overpower limit, departure from
nucleate boiling ratio (DNBR) limits, heat flux hot channel factor
(FQ), nuclear enthalpy rise hot channel factor (FDH), loss of
coolant accident peak cladding temperature (LOCA PCT), peak local
power density, or any other margin of safety. The radiological dose
consequence acceptance criteria listed in the Standard Review Plan
will continue to be met.
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: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and 2, Louisa County, Virginia
Date of amendment request: December 13, 2002.
Description of amendment request: The proposed amendments will
extend the Completion Time of Technical Specification (TS) 3.8.7,
Inverters-Operating, Required Action A.1, from 24 hours to 14 days for
an inoperable inverter on either Train H or Train J.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
1. The proposed license amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change to extend the Completion Time for an
inoperable inverter from 24 hours to 14 days does not alter any
plant equipment or operating practices in such a manner that the
probability of an accident is increased. In addition, this proposed
change will not alter assumptions relative to the mitigation of an
accident or transient event.
The licensee performed an evaluation to determine the risk
significance of the proposed change. This risk evaluation concluded
that the increases in annual core damage frequency (CDF) and large
early release frequency (LERF) associated with the proposed change
can be characterized as ``very small changes'' by Regulatory Guide
(RG) 1.174, ``An Approach for Using Probabilistic Risk Assessment in
Risk-Informed Decisions on Plant-Specific Changes to the Licensing
Basis.'' Additional evaluation by the licensee determined that the
incremental conditional core damage probability (ICCDP) and
incremental conditional large early release probability (ICLERP)
associated with the proposed change are within the acceptance
criteria in RG 1.177, ``An Approach for Plant-Specific, Risk-
Informed Decisionmaking: Technical Specifications.'' Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed license amendment 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.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change to extend the Completion Time for an
inoperable inverter has been evaluated for its effect on plant
safety. The licensee's risk-informed evaluation concluded that the
increases in annual CDF and LERF associated with the proposed change
can be characterized as ``very small changes'' by RG 1.174. The
[[Page 18290]]
ICCDP and ICLERP associated with the proposed change are within the
acceptance criteria in RG 1.177. Therefore, the proposed change does
not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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 Nuclear
Counsel, Dominion Nuclear Connecticut, Inc., Millstone Power Station,
Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford,
Connecticut 06385.
NRC Section Chief: John A. Nakoski.
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].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of application for amendment: November 27, 2002.
Brief description of amendment: This amendment deletes technical
specification (TS) 5.5.3, ``Post Accident Sampling,'' and thereby
eliminates the requirements to have and maintain the post accident
sampling system at the Clinton Power Station, Unit 1. The amendment
also addresses related changes to TS 5.5.2, ``Primary Coolant Sources
Outside Containment.''
Date of issuance: March 21, 2003.
Effective date: As of the date of issuance and shall be implemented
within 180 days.
Amendment No.: 155.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 21, 2003 (68 FR
2797).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 21, 2003.
No significant hazards consideration comments received: No.
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of application for amendment: April 10, 2002, as supplemented
February 12, 2003.
Brief description of amendment: The amendment deleted Technical
Specification 4.6.1.c, related to 24-month emergency diesel generator
surveillance, and relocated these requirements to the Updated Final
Safety Evaluation Report (UFSAR).
Date of issuance: April 3, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days, including the relocation of the emergency diesel
generator maintenance requirements of Technical Specification 4.6.1.c
to the Updated Final Safety Analysis Report (UFSAR), as was described
in the licensee's application dated April 10, 2002, and evaluated in
the NRC staff's safety evaluation dated April 3, 2003, and which
relocation shall be included in the next scheduled update of the UFSAR
pursuant to 10 CFR 50.71(e).
Amendment No.: 243.
Facility Operating License No. DPR-50: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 28, 2002 (67 FR
36926).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 3, 2003.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: July 24, 2002, as supplemented
February 21, 2003.
Brief Description of amendments: The amendments revise the
Technical Specifications Section 3.1.7, ``Standby Liquid Control (SLC)
System,'' to reflect modifications being made to the system as a result
of transition to the GE14 fuel design.
Date of issuance: March 25, 2003.
Effective date: March 25, 2003.
Amendment Nos.: 227 and 255.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the Technical Specifications and Appendix B, ``Additional
Conditions.''
Date of initial notice in Federal Register: August 20, 2002 (67 FR
53984).
The February 21, 2003, supplement contained clarifying information
only and did not change the initial no significant hazards
consideration determination or expand the scope of the initial Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 25, 2003.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket No. 50-324, Brunswick Steam
Electric Plant, Unit 2, Brunswick County, North Carolina
Date of amendment request: November 7, 2002, as supplemented
February 17, 2002.
Brief description of amendment: The amendment revises the Minimum
Critical Power Ratio (MCPR) Safety Limit contained in Technical
Specification 2.1.1.2 from 1.09 to 1.11
[[Page 18291]]
for two recirculation loop operation and from 1.10 to 1.13 for single
recirculation loop operation.
Date of issuance: March 25, 2003.
Effective date: March 25, 2003.
Amendment No.: 254.
Facility Operating License No. DPR-62: Amendment changes the
Technical Specifications.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75869). The February 17, 2003, supplement contained clarifying
information only and did not change the initial no significant hazards
consideration determination or expand the scope of the initial Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 25, 2003.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: August 28, 2002, as supplemented
November 21, 2002.
Brief description of amendment: This amendment revises the
Technical Specifications (TS) by adding Topical Report EMF-2328 (P)(A),
``PWR Small Break LOCA Evaluation Model, S-RELAP5 Based'' as reference
in the TS to allow the licensee to update the methodologies that are
used for safety analyses for the Shearon Harris Nuclear Power Plant,
Unit 1. The amendment also relocates referenced methodologies within TS
6.9.1.6.2 to group mechanical design methodologies together.
Date of issuance: March 28, 2003.
Effective date: March 28, 2003.
Amendment No.: 114.
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: October 15, 2002 (67 FR
63691). The November 21, 2002, supplement contained clarifying
information only and did not change the initial proposed no significant
hazards consideration determination or expand the scope of the initial
application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 28, 2003.
No significant hazards consideration comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: May 23, 2002, as supplemented
December 20, 2002, and February 27, 2003.
Brief description of amendment: The amendment revises the Fermi 2
Technical Specifications (TSs) to allow a one-time deferral of the Type
A primary containment integrated leak rate test. Specifically, TS
5.5.12, ``Primary Containment Leakage Rate Testing Program,'' would be
revised to extend the current interval for performing the containment
Type A test to 15 years.
Date of issuance: March 27, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 153.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: June 25, 2002 (67 FR
42817).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 27, 2003.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: November 6, 2001, as
supplemented on December 27, 2001, and July 15, August 6, and October
29, 2002.
Brief description of amendment: The amendment revises the Technical
Specifications (TSs) associated with the spent fuel pool (SFP).
Specifically, the amendment increases the allowable nominal average
fuel assembly enrichment from 4.5 weight percent (w/o) Uranium-235 (U-
235) to 4.85 w/o U-235 for all regions of the SFP, the new fuel storage
racks (dry), and the reactor core; allows fuel to be located in the 40
storage cells in Region B of the SFP that are currently empty and
blocked; credits SFP soluble boron for reactivity control during normal
conditions; and reduces the Boraflex reactivity credit in Regions A and
B of the SFP.
Date of issuance: April 1, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 274.
Facility Operating License No. DPR-65: This amendment revised the
TSs.
Date of initial notice in Federal Register: February 19, 2002 (67
FR 7414). The supplement dated December 27, 2001, provided a revision
to the licensee's analysis of the issue of no significant hazards
consideration, as originally provided in the November 6, 2001,
application. The supplements dated July 15, August 6, and October 29,
2002, provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 1, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: June 3, 2002, as supplemented on
January 23, 2003.
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.4.9, ``Pressurizer,'' to increase the pressurizer
water level limit when the plant is in MODE 3 (Hot Standby). The
pressurizer water level limit for MODES 1 and 2 (Power Operation and
Startup) remains unchanged. The amendment also revises TS 3.8.4, ``DC
Sources--Operating,'' to remove the notes that refer to the one-time
amendment allowing the online replacement of station batteries 31 and
32. The notes were no longer applicable since the batteries have been
replaced.
Date of issuance: March 25, 2003.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 216.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 9, 2002 (67 FR
45566).
The January 23 letter provided clarifying information that did not
enlarge the scope of the original Federal Register notice or change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 25, 2003.
No significant hazards consideration comments received: No.
[[Page 18292]]
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: December 4, 2002, which replaces
the original applications dated May 1, 2002.
Brief description of amendment: The proposed amendment would change
the Pilgrim Nuclear Power Station Technical Specification (TS) Figures
3.6.1, 3.6.2, and 3.6.3 to extend the applicability of the current
reactor pressure vessel pressure-temperature (P-T) curves through the
end of Operating Cycle (OC) 16. The current P-T curves were approved
for use in License Amendment 190, dated April 13, 2001, and are limited
to use through the end of OC 14. The proposed change would delete the
20 and 32 Effective Full Power Year curves and replace the wording of
the title blocks to allow use through the end of OC 16.
Date of issuance: March 28, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 197.
Facility Operating License No. DPR-35: This amendment revised the
TS.
Date of initial notice in Federal Register: February 18, 2003 (68
FR 7816).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 28, 2003.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: August 19, 2002, as supplemented by
letter dated December 19, 2002.
Brief description of amendment: The amendment revises the Technical
Specifications by: (1) Modifying the wording of the current
Surveillance Requirement (SR) 4.0.1 and SR 4.0.3 to be consistent with
NUREG-1431, Revision 2, Improved Standard Technical Specifications
(ISTS) wording for SR 3.0.1 and SR 3.0.3; and (2) modifying the ISTS
wording, adopted in Item (1), above, for SR 4.0.3 to extend the delay
period, before entering a Limiting Condition for Operation, following a
missed surveillance. The delay period is extended from the current
limit of up to 24 hours ``* * * when the allowable outage time limits
of the ACTION requirements are less than 24 hours'' to ``* * * up to 24
hours or up to the limit of the specified surveillance interval,
whichever is greater.'' In addition, the following requirement is added
to SR 4.0.3: ``A risk evaluation shall be performed for any
Surveillance delayed greater than 24 hours and the risk impact shall be
managed.''
Date of issuance: March 21, 2003.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 187.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications and Surveillance Requirements.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5670).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 21, 2003.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: December 20, 2002.
Brief description of amendment: This amendment approves several
administrative changes to the Waterford Steam Electric Station, Unit 3
Technical Specifications (TSs) to revise, correct, or clarify certain
titles, page numbers, and heading information. It also revises
personnel and committee titles that have been changed, revises
administrative reporting requirements to conform to 10 CFR 50.4, and
deletes redundant or unnecessary requirements from TSs 5.4, 6.6, and
6.7.
Date of issuance: April 3, 2003.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 188.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5673).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 3, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: August 16, 2002.
Brief description of amendments: The amendments modify the Unit 3
allowable value Technical Specification, and the Units 2 and 3
surveillance requirements Technical Specification for the reactor
protection system scram discharge volume water level-high function.
Date of issuance: April 3, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 198/191.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68737).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 3, 2003.
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: November 27, 2002.
Brief description of amendments: These amendments delete Technical
Specification (TS) 5.5.3, ``Post Accident Sampling,'' and thereby
eliminate the requirements to have and maintain the post accident
sampling system at the LaSalle County Station, Units 1 and 2. The
amendments also address related changes to TS 5.5.2, ``Primary Coolant
Sources Outside Containment.''
Date of issuance: March 21, 2003.
Effective date: As of the date of issuance and shall be implemented
within 180 days.
Amendment Nos.: 158/144.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 21, 2003 (68 FR
2802).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 21, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: August 22, 2002.
Brief description of amendments: The amendments modify the required
surveillance interval from monthly to quarterly for calibration of the
trip units associated with the instrumentation channels of the
Anticipated Transient Without Scram-Recirculation Pump Trip system.
Date of issuance: April 1, 2003.
[[Page 18293]]
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 213 and 207.
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 1, 2002 (67 FR
61682). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 1, 2003.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit 2, Berrien County, Michigan
Date of application for amendment: October 16, 2002, as
supplemented January 28, 2003.
Brief description of amendment: The amendment would revise the
Technical Specification values for the 4 kilovolt degraded-voltage and
loss-of-voltage relays.
Date of issuance: March 26, 2003.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 256.
Facility Operating License No. DPR-74: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68739).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 26, 2003.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station, Unit No. 1, Oswego County, New York
Date of application for amendment: December 19, 2002.
Brief description of amendment: The amendment revised the Technical
Specifications to add the definition of shutdown margin (SDM),
incorporate new, more restrictive SDM limits, add the associated
limiting condition for operation actions and completion times for each
applicable operating condition if the SDM is not met, and add
surveillance requirements for verifying SDM.
Date of issuance: March 27, 2003.
Effective date: March 27, 2003.
Amendment No.: 180.
Facility Operating License No. DPR-63: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 21, 2003 (68 FR
2806).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 27, 2003.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of application for amendment: March 29, 2002, as supplemented
by letter dated January 24, 2003.
Brief description of amendment: The amendment changes the
surveillance requirement of TS 5.5.12, ``Primary Containment Leakage
Rate Testing Program,'' to allow a one-time 5-year extension to the 10-
year interval for performing the next Type A containment integrated
leakage rate test (ILRT). The change allows ILRT testing within 15
years from the last ILRT, which was performed in September 1993.
Date of issuance: March 21, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 249.
Facility Operating License No. DPR 49: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 30, 2002 (67 FR
21291).
The supplemental letter contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 21, 2003.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: April 22, 2002, as supplemented
October 25, 2002, January 23, and February 12, 2003.
Brief description of amendment: The amendment changes TS
Surveillance Requirement 4.7.A.2.b, ``Primary Containment Integrity,''
to allow a one-time, 5-year extension to the 10-year interval for
performing the next Type A containment integrated leakage rate test
(ILRT). The change allows ILRT testing within 15 years from the last
ILRT, which was performed in March 1993.
Date of issuance: March 31, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 134.
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 3, 2002 (67
FR 56324).
The October 25, 2002, January 23, and February 12, 2003,
supplements provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the Nuclear Regulatory Commission staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 31, 2003.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: July 25, 2002, as supplemented
on October 21, 2002.
Brief description of amendment: The amendment would modify
Technical Specification (TS) requirements for missed surveillance tests
in TS 4.0.3 using the Consolidated Line Item Improvement Program,
modify TS 4.0.1 to be consistent with the Standard Technical
Specifications (STS), and incorporate a TS Bases Control Program in
Section 6.0 in accordance with the STS.
Date of issuance: March 31, 2003.
Effective date: March 31, 2003.
Amendment No.: 145.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75883)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 31, 2003.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: April 4, 2002, as supplemented by
letter dated January 9, 2003.
Brief Description of amendments: The amendments revise Technical
[[Page 18294]]
Specifications 5.5.17, ``Containment Leakage Rate Testing Program,'' to
reflect a one-time deferral of the Type A Containment Integrated Leak
Rate Test (ILRT). The 10-year interval between ILRTs is to be extended
to 15 years from the previous ILRTs that were completed in March 1994
for Unit 1 and March 1995 for Unit 2.
Date of issuance: March 21, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 159/150.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68743).
The supplement, dated January 9, 2003, provided clarifying
information that did not change the scope of the April 4, 2002,
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 March 21, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant,
Unit 2, Hamilton County, Tennessee
Date of application for amendment: November 15, 2002, as
supplemented by letters dated February 19, 2003, and February 26, 2003.
Description of amendment: This one-time condition establishes
special provisions and requirements for safe operation of Unit 2 while
heavy load lifts are performed during the Unit 1 steam generator
replacement project. The provisions for heavy load lifts are described
in Topical Report 24370-TR-C-002, which was previously submitted on
April 15, 2002, for NRC review and approval. The topical report
contains prerequisite actions for heavy load movement, active
monitoring during heavy load movement, and compensatory measures in
response to the unlikely event of a heavy load drop.
Date of issuance: March 26, 2003.
Effective date: As of the date of issuance and shall be implemented
within 45 days of issuance.
Amendment No.: 273.
Facility Operating License No. DPR-79: Amendment revises the
Operating License.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75885).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 26, 2003.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 7th day of April, 2003.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 03-9026 Filed 4-14-03; 8:45 am]
BILLING CODE 7590-01-P