[Federal Register Volume 68, Number 92 (Tuesday, May 13, 2003)]
[Notices]
[Pages 25648-25664]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-11697]
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NUCLEAR REGULATORY COMMISSION
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations; Biweekly Notice
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 18, 2003, through May 1, 2003. The
last biweekly notice was published on April 29, 2003 (68 FR 22744).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
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
[[Page 25649]]
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 June 12, 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 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,
[[Page 25650]]
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].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: April 2, 2001, as supplemented by
letters dated January 15, August 23, 2002, and March 28, 2003.
Description of amendment request: The proposed amendment would add
operational restrictions when the inclined fuel transfer system (IFTS)
blind flange is removed during Modes 1, ``Power Operation,'' 2,
``Startup,'' or 3, ``Hot Shutdown.'' The proposed changes would (1)
include a limitation on the duration that the IFTS blind flange can be
removed while primary containment integrity is required, (2) include a
limitation on the duration that the IFTS blind flange can remain in the
unbolted configuration, (3) specify the need to install the steam dryer
pool to reactor cavity pool gate prior to opening the blind flange, and
(4) provide the flexibility to remove the IFTS blind flange for other
than maintenance and testing purposes only.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes allow operation of the IFTS while primary
containment operability is required. The proposed changes result in
a change to the primary containment boundary. A loss of primary
containment integrity is not an accident initiator. The proposed
changes do not involve any modifications to plant systems or design
parameters or conditions that contribute to the initiation of any
accidents previously evaluated. Therefore, the proposed changes do
not increase the probability of any accident previously evaluated.
The proposed changes potentially affect the allowable leakage of
the containment structure which is designed to mitigate the
consequences of a loss-of-coolant accident (LOCA). The function of
the primary containment is to maintain functional integrity during
and following the peak transient pressures and temperatures that
result from any LOCA. The primary containment is designed to limit
fission product leakage following the design basis LOCA. Because the
proposed changes do not alter the plant design, only the extent of
the boundaries that provide primary containment isolation for the
IFTS penetration, the proposed changes do not result in an increase
in primary containment leakage. In addition, a time limit for IFTS
blind flange removal of 40 days per cycle and a 12 hour limit for
the unbolted configuration of the IFTS flange have been established
as conservative measures to limit the associated risk to the
containment boundary for all accident conditions. Once the blind
flange is removed the IFTS transfer tube and its appurtenances
become part of the primary containment boundary. As part of the
primary containment boundary these subject components would be
exposed to LOCA pressures. While these components have not been
fabricated or installed to meet the acceptance criteria for a
containment penetration, they have been built to withstand the
rigors of a commercial nuclear application. This includes, but is
not limited to, consideration of adequate seismic support, inertial
forces imparted to the fuel, appropriate cooling and shielding for
the spent nuclear fuel, integrity of the fluid system pressure
boundary, and a safety analysis, including a failure modes and
effects evaluation which assumes that credible events and credible
combinations of events have been considered and mitigated against by
either a fail safe design or redundancy. They are judged to be an
acceptable barrier to prevent the uncontrolled release of post-
accident fission products for the purposes of this amendment
request.
Further, it has been shown that the largest potential leakage
pathway, the IFTS transfer tube itself, would remain sealed by the
depth of water required by the proposed [technical specification] TS
change to be maintained in the fuel building fuel transfer pool. The
transfer tube drain line constitutes the other possible leakage
pathway, and will be required to be capable of being isolated via
administrative control of the manual isolation valve in the drain
line. Additionally, due to the physical relationships of the
buildings and components involved, any leakage from either of these
pathways is fully contained within the boundaries of the secondary
containment and would be filtered by the Standby Gas Treatment
System prior to release to the environment.
Leakage from the containment upper pool through the open IFTS
transfer tube could potentially result in the excessive loss of
water from the volume intended to provide post-LOCA makeup water to
the suppression pool. The upper pool dump volume is maintained by
requiring the installation of the steam dryer pool to reactor cavity
pool gate with the seal inflated and a backup air supply provided.
Maintaining the upper pool dump volume ensures proper suppression
pool level can be achieved following a LOCA which provides for long-
term steam condensation.
Based on the above, the proposed changes do not increase the
consequences of an accident previously evaluated.
In summary, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not involve a change to the plant design
or operation except for when IFTS is operated. As a result, the
proposed changes do not affect any of the parameters or conditions
that could contribute to the initiation of any accidents. No new
accident modes or equipment failure modes are created by these
changes. Extending the primary containment boundary to include
portions of the IFTS has no influence on, nor does it contribute to
the possibility of a new or different kind of accident or
malfunction from those previously evaluated. Furthermore, operation
of IFTS is unrelated to the operation of the reactor. There is no
mishap in the process that can lead or contribute to the possibility
of losing any coolant in the reactor or introducing the chance for
positive or negative reactivity or other accidents different from
and not bounded by those previously evaluated. Therefore, these
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed changes only affect the extent of a portion of the
primary containment boundary. The time that the IFTS is in the
seismically indeterminate configuration with the flange unbolted
will be limited to 12 hours per operating cycle. The time the IFTS
blind flange will be removed will be limited to 40 days per
operating cycle. These restrictions will limit the risk from the
potential leakage through the primary containment boundary. Having
IFTS in operation does not affect the reliability of equipment used
for core cooling. In addition, precautions will be taken to
administratively control the IFTS transfer tube drain path so that
the proposed change will not increase the probability that an
increase in leakage from the primary containment to the secondary
containment could occur. Precautions will also be taken to ensure
that the steam dryer pool to reactor cavity pool gate is installed
prior to removing the IFTS flange when primary containment is
required to be operable. Installation of this gate will ensure that
an adequate containment upper pool dump volume is maintained to
support post-LOCA suppression pool makeup water volume requirements.
The margin of safety that has the potential of being impacted by
the proposed changes involve the offsite dose consequences of
postulated accidents which are directly related to containment
leakage rate. The containment isolation system is designed to limit
leakage to La which is defined by the
[[Page 25651]]
[Clinton Power Station] CPS TS to be 0.65% of primary containment
air weight per day at the design basis LOCA maximum peak containment
pressure (i.e., Pa). The limitation on containment
leakage rate is designed to ensure that total leakage volume will
not exceed the volume assumed in the accident analyses at
Pa. The margin of safety for the offsite dose
consequences of postulated accidents directly related to the
containment leakage rate is maintained by meeting the La
acceptance criteria during operation. The La value is not
being modified by this proposed TS change. The IFTS will continue to
provide an acceptable barrier to prevent unacceptable containment
leakage during a LOCA, and therefore these changes will not create a
situation causing the containment leakage rate acceptance criteria
to be violated.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Edward J. Cullen, Deputy General Counsel
Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: March 28, 2003.
Description of amendments request: The amendment would remove the
post-accident hydrogen monitoring and control requirements from the
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 Technical
Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change to the Technical Specifications has been
evaluated against the standards in 10 CFR 50.92. The proposed
amendment revises Technical Specification 3.3.10, Post-Accident
Monitoring Instrumentation, and Technical Specification Table
3.3.10-1, Post-Accident Monitoring Instrumentation to delete
references to the containment hydrogen analyzers. Additionally, the
proposed amendment will delete Technical Specification 3.6.7,
Hydrogen Recombiners. The proposed change has been determined to not
involve a significant hazards consideration, in that operation of
the facility in accordance with the proposed amendments:
1. Would not involve a significant increase in the probability
or consequences of any accident previously evaluated.
Components used in the control of hydrogen in the Containment
(consisting of hydrogen recombiners, a hydrogen vent, and hydrogen
detectors) are not considered accident initiators. Therefore, this
change does not increase the probability of an accident previously
evaluated.
The purpose of the Hydrogen Control System is to ensure that
hydrogen concentration is maintained below 4.0 volume percent so
that Containment integrity is not challenged following a design
basis loss-of-coolant accident (LOCA). The Calvert Cliffs Nuclear
Power Plant Individual Plant Examination analyzed the probability of
Containment failure under a variety of conditions. This proposed
amendment does not alter the conclusions or assumptions of the
Individual Plant Examination. The Calvert Cliffs Nuclear Power Plant
Containment provides a safety margin against hydrogen burn following
a design basis accident, such that the Containment will not fail
even without hydrogen control equipment. Therefore, this change does
not increase the consequences of accidents previously evaluated.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. Would not create the possibility of a new or different [kind]
of accident from any accident previously evaluated.
The proposed change does not change the configuration of the
plant beyond the Hydrogen Control System. Hydrogen generation
following a design basis LOCA has been evaluated. Deletion of the
Hydrogen Control System from the plant design basis and Technical
Specifications does not alter the generation of hydrogen post-LOCA.
Therefore, this change does not create the possibility of a new
or different [kind] of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in [a] margin of
safety.
The margin of safety in this case is the ability of Containment
to withstand a pressure increase caused by the deflagration of
hydrogen in the Containment. Industry experience and experimentation
has shown that large, dry, well-ventilated Containments such as
those at Calvert Cliffs can withstand pressures generated by
ignition of hydrogen resulting from a LOCA. The Calvert Cliffs
Nuclear Power Plant Containment provides a safety margin against
hydrogen burn following a design basis accident, such that the
Containment will not fail even without hydrogen control equipment.
Therefore, this change does not significantly reduce [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
amendments request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Richard J. Laufer.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,
Unit No. 1, Pope County, Arkansas
Date of amendment request: April 2, 2003.
Description of amendment request: The proposed amendment would
change the spent fuel pool loading restrictions by redefining the
regions, inserting Metamic[reg] poison panels in a portion of the spent
fuel pool, and increasing the minimum boron concentration.
Basis for 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 three fuel handling accidents described below can be
postulated to increase reactivity. However, for these accident
conditions, the double contingency principle of ANS [American
Nuclear Society] N16.1-1975 is applied. This states that it is
unnecessary to assume two unlikely, independent, concurrent events
to ensure protection against a criticality accident. Thus, for
accident conditions, the presence of soluble boron in the storage
pool water can be assumed as a realistic initial condition since its
absence would be a second unlikely event.
Three types of drop accidents have been considered: a vertical
drop accident, a horizontal drop accident, and an inadvertent drop
of an assembly between the outside periphery of the rack and the
pool wall:
[sbull] A vertical drop directly upon a cell will cause damage
to the racks in the active fuel region. The current 1600 ppm soluble
boron concentration TS limit will ensure that Keff does
not exceed 0.95.
[sbull] A fuel assembly dropped on top of the rack horizontally
will not deform the rack structure such that criticality assumptions
are invalidated. The rack structure is such that an assembly
positioned horizontally on top of the rack results in a separation
distance from the upper end of the active fuel region of the stored
assemblies. This distance is sufficient to preclude interaction
between the dropped assembly and the stored fuel.
[sbull] An inadvertent drop of an assembly between the outside
periphery of the rack and the pool wall is bounded by the worst case
fuel misplacement accident condition.
The fuel assembly misplacement accident was considered for all
storage configurations. An assembly with high reactivity is assumed
[[Page 25652]]
to be placed in a storage location which requires restricted storage
based on initial U-235 loading, cooling time, and burnup. The
presence of boron in the pool water assumed in the analysis has been
shown to offset the worst case reactivity effect of a misplaced fuel
assembly for any configuration. This boron requirement is less than
the 1600 ppm currently required by the ANO-1 TS. Thus, a five
percent subcriticality margin can be easily met for postulated
accidents, since any reactivity increase will be much less than the
negative worth of the dissolved boron.
For fuel storage applications, water is usually present. An
``optimum moderation'' accident is not a concern in spent fuel pool
storage racks because the rack design prevents the preferential
reduction of water density between the cells of a rack (e.g.,
boiling between cells). An ``optimum moderation'' accident in the
new fuel pit was previously evaluated and the conclusions of that
evaluation have not changed as a result of the fuel enrichment.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes will define a portion of the current Region
2 as Region 3. The new region will contain Metamic[reg] poison panel
inserts and will allow unrestricted storage of fuel assemblies with
various enrichments and burnup. To support the proposed change, new
criticality analyses have been performed. The analyses resulted in
new loading restrictions in Region 1 and Region 2. The presence of
boron in the pool water assumed in the analysis is less than the
1600 ppm currently required by the ANO-1 TSs.
Thus, a five percent subcriticality margin can be easily met for
postulated accidents, since any reactivity increase will be much
less than the negative worth of the dissolved boron.
No new or different types of fuel assembly drop scenarios are
created by the proposed change. During the installation of the
Metamic[reg] panels, the possible drop of a panel is bounded by the
current fuel assembly drop analysis. No new or different fuel
assembly misplacement accidents will be created. Administrative
controls currently exist to assist in assuring fuel misplacement
does not occur.
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.
With the presence of a nominal boron concentration, the SFP
storage racks will be designed to assure that fuel assemblies of
less than or equal to five weight percent U-235 enrichment when
loaded in accordance with the proposed loading restrictions will be
maintained within a subcritical array with a five percent
subcritical margin (95% probability at the 95% confidence level).
This has been verified by criticality analyses.
Credit for soluble boron in the SFP water is permitted under
accident conditions. The proposed modification that will allow
insertion of Metamic[reg] poison panels does not result in the
potential of any new misplacement scenarios. Criticality analyses
have been performed to determine the required boron concentration
that would ensure the maximum Keff does not exceed 0.95.
The ANO-1 TS for the minimum SFP boron concentration is greater than
that required to ensure Keff does not exceed 0.95.
Therefore, the margin of safety currently defined by taking credit
for soluble boron will be maintained.
The structural analysis of the spent fuel racks, along with the
evaluation of the SFP structure, showed that the integrity of these
structures will be maintained with the addition of the poison
inserts. The structural requirements were shown to be satisfied, so
the safety margins were maintained.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: March 19, 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 (TMI-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 U.S. 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 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 Federal Register on March 20, 2002 (67 FR 13027).
The licensee affirmed the applicability of the following NSHC
determination in its application dated March 19, 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
[[Page 25653]]
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: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: April 3, 2003.
Description of amendment request: The proposed changes will revise
the Updated Final Safety Analysis Report to change the Reactor Vessel
Material Surveillance Program. The change will reflect participation in
the Boiling Water Reactor Vessel and Internals Project (BWRVIP)
Integrated Surveillance Program (ISP).
Basis for proposed no significant hazards consideration
determination: As required by Section 50.91(a) of Title 10 of the Code
of Federal Regulations (10 CFR ), 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.
Changes in the fracture toughness properties of reactor vessel
beltline materials, resulting from the neutron irradiation and the
thermal environment, are monitored by a surveillance program in
compliance with the requirements of 10CFR50, Appendix H. The
proposed change implements an integrated surveillance program that
has been evaluated by the NRC [U.S. Nuclear Regulatory Commission]
staff as meeting the requirements of paragraph III.C of Appendix H
to 10 CFR 50. The BWRVIP's ISP surveillance material selection
process adequately ensures that materials in the program effectively
provide meaningful information to monitor changes in fracture
toughness for GGNS [Grand Gulf Nuclear Station, Unit 1, or Grand
Gulf] RPV [Reactor Pressure Vessel] materials. In addition, the ISP
program requires participants to acquire and evaluate relevant ISP
test data from the program which may affect RPV integrity
evaluations in a timely manner. One advantage of participating in
the BWRVIP ISP is that surveillance test data applicable to the
Grand Gulf RPV will be available sooner than under the current plant
specific program.
The proposed change will not affect current RPV performance and
will not cause the RPV or interfacing systems to be operated outside
of their design or testing limits. The proposed change will not
alter any assumptions previously made in evaluating the radiological
consequences of accidents.
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 change does not affect the design, function, reliability, or
operation of any plant structure, system or component. The purpose
of the reactor vessel material surveillance program is to monitor
neutron embrittlement and thermal environment effects in order to
predict the behavioral characteristics of materials of pressure
retaining components of the reactor coolant pressure boundary and to
ensure that reactor vessel fracture toughness and integrity
requirements are not violated. The ISP is an approved alternate
monitoring program that meets the regulatory requirements in
Appendix H to 10 CFR 50. As an acceptable alternate monitoring
program, the ISP cannot create a new failure mode involving the
possibility of a new or different kind of accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from that previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The reactor material surveillance program required by 10 CFR 50,
Appendix H, is designed to ensure that adequate margins of safety
are provided for the reactor coolant pressure boundary during any
condition of normal operation, including anticipated operational
occurrences and hydrostatic tests. Monitoring changes in the
fracture toughness of reactor vessel materials ensures that material
changes due to radiation embrittlement are adequately considered for
safe reactor operations. Paragraph lll.C of Appendix H to 10 CFR 50
delineates the regulatory requirements for an ISP. The BWRVIP ISP
meets these requirements and has been approved by the NRC.
One of the uses of the material surveillance data obtained
through the proposed ISP is to ensure the reactor coolant system P/T
[Pressure/Temperature] limits established by the Technical
Specifications are conservative. The material surveillance data
obtained through the proposed Integrated Surveillance Program will
provide new information that will be evaluated to ensure that the P/
T limits are conservative. In addition, a neutron fluence
calculation methodology which has been approved by the NRC staff and
is consistent with the attributes identified in U.S. Nuclear
Regulatory Commission Regulatory Guide 1.190, ``Calculational and
Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,''
will be used for the determination of reactor vessel and
surveillance capsule neutron fluence values to ensure quality of the
method and compatibility between ISP results.
[[Page 25654]]
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: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units
1 and 2, Rock Island County, Illinois
Date of application for amendment request: February 14, 2003.
Description of amendment request: The proposed amendments would
relax the Technical Specifications (TSs) surveillance requirement (SR)
for reactor instrumentation line excess flow check valves (EFCVs).
Currently, TSs require testing of each reactor instrumentation line
EFCV on a 24-month frequency. The proposed TS SR would require that a
representative sample of reactor instrumentation line EFCVs be tested
every 24 months, such that each EFCV will be tested nominally once
every 10 years.
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 change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The current Technical Specification (TS) Surveillance
Requirement (SR) frequency requires each reactor instrumentation
line excess flow check valve (EFCV) to be tested every 24 months.
The EFCVs at Dresden Nuclear Power Station (DNPS) and Quad Cities
Nuclear Power Station (QCNPS) are designed to remain open during
normal operation, but will close automatically in the event of an
instrument line break downstream of the valve. The proposed change
allows a reduced number of reactor instrumentation line EFCVs to be
tested every 24 months. Industry operating experience demonstrates a
high level of reliability for these EFCVs. A failure of an EFCV to
isolate cannot initiate previously evaluated accidents (i.e., a
break in a reactor coolant pressure boundary (RCPB) instrument line
outside containment). Therefore, there is no increase in the
probability of an accident as a result of this proposed change.
The postulated break of an instrument line connected to the RCPB
is discussed and evaluated in the Updated Final Safety Analysis
Reports (UFSARs) for DNPS and QCNPS. The integrity and functional
performance of the secondary containment and standby gas treatment
system are not impaired by this event, and the calculated potential
offsite exposures are below the guidelines of 10 CFR 100, ``Reactor
Site Criteria.'' The NRC approved General Electric Nuclear Energy
Licensing Topical Report, NEDO-32977-A, ``Excess Flow Check Valve
Testing Relaxation,'' discusses through operating experience that
there is a high degree of reliability with the EFCVs and that there
are little radiological consequences resulting from an EFCV failure.
The radiological consequences for an instrument line break do not
credit the EFCVs for isolating the break. Therefore, the
consequences of an instrument line break are not impacted by the
proposed level of testing. Based on the above, the proposed TS
change does not involve a significant increase in the consequences
of an accident previously evaluated.
In summary, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change allows a reduced number of reactor
instrumentation line EFCVs to be tested every 24 months. No other
changes in requirements are being proposed. Industry operating
experience as documented in NEDO-32977-A, provides supporting
evidence that the reduced testing will not affect the high
reliability of these valves. The potential failure of an EFCV to
isolate as a result of the proposed reduction in testing is bounded
by the evaluation of an instrument line break described in the
UFSARs for DNPS and QCNPS. The proposed changes do not physically
alter the plant and will not alter the operation of structures,
systems, and components as described in the UFSARs. Therefore, a new
or different kind of accident from any accident previously evaluated
will not be created.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
The consequences of an unisolable rupture of a RCPB instrument
line outside containment has been previously evaluated in the UFSARs
for DNPS and QCNPS. That evaluation assumed a continuous discharge
of reactor coolant for the duration of the detection and cooldown
sequence (i.e., no credit was assumed for isolating the break by the
associated EFCV in the ruptured instrument line). Since a continuous
discharge was assumed in this evaluation, any potential failure of
the associated EFCV to isolate postulated by the reduced testing
frequency is bounded. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Vice President,
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way,
Kennett Square, PA 19348.
NRC Section Chief: Anthony J. Mendiola.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: March 31, 2003.
Description of amendment request: The proposed amendments would
revise Appendix A, Technical Specifications (TS), of Facility Operating
License Nos. NPF-11 and NPF-18. Specifically, the proposed change will
increase the upper limit associated with TS Table 3.3.5.1-1,
``Emergency Core Cooling System Instrumentation,'' Function 3.e, ``HPCS
System Flow Rate--Low (Bypass),'' Allowable Value from less than or
equal to (<=) 1704 gallons per minute (gpm) to <= 2194 gpm. The
proposed change increases the Allowable Value band to account for
instrumentation deadband, as-left setting tolerances and setpoint
drift, and resolves historical difficulties during calibration. The
current Allowable Value was initially provided in the LaSalle County
Station TS during conversion to Improved Technical Specifications (ITS)
format. This value was based on vendor supplied data and believed at
the time to adequately account for these parameters. The upper
Allowable Value limit is being increased based on historical
performance data for the High Pressure Core Spray (HPCS) system flow
switches. The increase in the allowed bypass flow rate does not affect
the capability of the HPCS system in performing its intended safety
function.
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 probability or consequences of an accident previously evaluated.
The proposed change to LaSalle County Station Technical
Specifications (TS) Table 3.3.5.1-1, ``Emergency Core Cooling System
[[Page 25655]]
Instrumentation,'' Function 3.e, ``HPCS System Flow Rate--Low
(Bypass),'' request an increase in the Allowable Value from less
than or equal to <= 1704 gpm to <= 2194 gpm. The operation of High
Pressure Core Spray (HPCS) System Flow Rate--Low (Bypass) function
is not a precursor to any accident previously evaluated. Thus, the
proposed change does not have any effect on the probability of an
accident previously evaluated.
The LaSalle County Station Emergency Core Cooling Systems (ECCS)
are designed, in conjunction with the primary and secondary
containments, to limit the release of radioactive material to the
environment following a loss of coolant accident (LOCA). The ECCS
uses two independent methods, flooding and spraying, to cool the
reactor core following a LOCA. The HPCS is one of the core spray
systems. The evaluation of the proposed change concluded that the
HPCS will operate as assumed in accidents previously evaluated.
Thus, the radiological consequences of any accident previously
evaluated are not increased.
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.
The proposed change does not affect the control parameters
governing unit operation and does not introduce any new equipment,
modes of system operation or failure mechanisms. Calculations have
been performed which evaluated the performance of the HPCS system
without the closure of the minimum flow bypass valve. The
calculations determined that the Unit 1 and Unit 2 HPCS pump
capacity with the minimum flow bypass valve open will support HPCS
System injection flow into the reactor pressure vessel (RPV) over
the full range of RPV pressures above the requirements for HPCS in
the Loss of Coolant Accident (LOCA) analyses.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The HPCS System Flow Rate--Low (Bypass) Function is one of the
inputs to the logic that controls the opening and closing of the
minimum flow bypass valve. The current Allowable Values for this
function are greater than or equal to (=)1380 gpm and
<=1704 gpm. The lower Allowable Value limit (i.e., 1380 gpm) ensures
that the minimum flow bypass valve opens when pump flow is too low
for adequate cooling of the pump while the pump is operating. This
limit is not affected by the proposed change.
The upper Allowable Value limit (i.e., 1704 gpm) ensures that
the minimum flow bypass valve automatically closes to allow maximum
flow to the RPV spray sparger. The proposed change increases the
value to <= 2194 gpm. LaSalle County Station has evaluated the
effect of this change and concluded the following:
[sbull] The proposed change to increase the upper Allowable
Value limit from <= 1704 gpm to <= 2194 gpm will provide further
assurance that the minimum flow bypass valve remains full open until
the HPCS pump flow to the RPV spray sparger is sufficient to prevent
overheating of the pump, and
[sbull] The upper Allowable Value ensures that the HPCS minimum
flow bypass valve closes to allow maximum flow to the RPV spray
sparger. The proposed change will delay the initiation of valve
closure from <= 1704 gpm to <= 2194 gpm. The calculations determined
that the Unit 1 and Unit 2 HPCS pump capacity with the minimum flow
bypass valve open will support HPCS system injection flow into the
RPV over the full range of RPV pressures above the requirements for
HPCS in the Loss of Coolant Accident (LOCA) analysis up to the
maximum assumed injection flow of 5400 gpm. The margin to the flow
requirements of the LOCA analysis varies from approximately 200 gpm
at very low RPV pressures to greater than 1000 gpm at higher RPV
pressures. Since the HPCS system injection flow requirement to the
RPV spray sparger assumed in the LOCA analysis is met with the
minimum flow bypass valve open, the LOCA analysis results are not
adversely affected by increasing the value of flow when the minimum
flow bypass valve starts to close. Although the calculations show
that closure of the HPCS minimum flow bypass valve is not necessary
to meet the HPCS system injection flow requirements assumed in the
LOCA analyses, LaSalle County Station has chosen to retain the upper
Allowable Value in the TS to provide additional margin to the
assumed injection flow of the analyses.
Thus, increasing the TS upper Allowable Value limit for the HPCS
System Flow Rate--Low (Bypass) Function from <= 1704 gpm to <= 2194
gpm will not affect the capability of the HPCS system in performing
its intended safety function.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
Based upon the above, Exelon Generation Company concludes that
the proposed amendment presents no significant hazards consideration
under the standards set forth in 10 CFR 50.92(c), and, accordingly,
a finding of ``no significant hazards consideration'' is justified.
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.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: March 19, 2003.
Description of amendment request: The proposed amendment would
delete requirements from the Technical Specifications (TSs) and other
elements of the licensing bases related to the post-accident sampling
system (PASS) at the Monticello Nuclear Generating Plant. 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 TSs 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 proposed changes are based on 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 on December 27, 2001 (66 FR 66949),
on possible amendments concerning TSTF-413. The notice included a model
safety evaluation and model no significant hazards consideration
determination, using the consolidated line-item improvement process.
The NRC staff subsequently issued a notice of availability of the
models for referencing in license amendment applications in the Federal
Register on March 20, 2002 (67 FR 13027). The licensee affirmed the
applicability of the following no significant hazards consideration
determination in its application dated March 19, 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:
[[Page 25656]]
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: 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. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: March 27, 2003.
Description of amendment request: The proposed amendment would
approve a selective scope application of an alternative source term
(AST) for fuel handling accidents (FHAs). Specifically, the amendments
would revise Technical Specification (TS) 3.9.3, ``Containment
Penetrations,'' to (1) change the Applicability statement to ``During
movement of recently irradiated fuel assemblies within containment,''
and (2) modify the Required Action for Condition A to eliminate the
requirement to suspend core alterations and add the requirement to
suspend movement of recently irradiated fuel assemblies within
containment if one or more containment penetrations are not in the
required status.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant increase in
the probability or consequences of any accident previously
evaluated.
Selective implementation of the Alternative Source Term (AST)
and those plant systems affected by implementing the proposed
changes to the TS are not accident initiators and cannot increase
the probability of an accident. The AST does not adversely affect
the design or operation of the facility in a manner that would
create an increase in the probability of an accident. Rather, the
AST is a methodology used to evaluate the dose consequences of a
postulated accident.
The fuel handling accident analysis has demonstrated that the
dose consequences of a postulated fuel handling accident remain
within the limits provided sufficient decay has occurred prior to
the movement of irradiated fuel without taking credit for certain
mitigation features such as ventilation filter systems and
containment closure. Irradiated fuel that has not undergone the
required decay period of 65 hours is defined as recently irradiated
fuel and the currently approved TS requirements are applicable when
this recently irradiated fuel is being handled.
This amendment does not alter the methodology or equipment used
directly in fuel handling operations. Neither ventilation filter
system (i.e., the containment purge or drumming area vent stack) is
used to actually handle fuel. Neither of these systems is an
accident initiator. Similarly, neither the equipment hatch,
personnel air locks, any other containment penetrations, nor any
component thereof is an accident initiator. No other accident
initiator is affected by the proposed changes.
The TEDE [total effective dose equivalent] doses from the
analysis supporting this amendment request have been compared to
equivalent TEDE doses estimated with the guidelines of RG
[Regulatory Guide] 1.183 [``Alternative Radiological Source Terms
for Evaluating Design Basis Accidents at Nuclear Power Reactors'']
Footnote 7. The new values are shown to be comparable to the results
of the previous analysis.
Based on the aforementioned reasons, the proposed amendment does
not involve a significant increase in the probability or
consequences of a FHA as previously analyzed.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a new or different kind
of accident from any accident previously evaluated.
The evaluation of the effects of the proposed changes indicates
that all design
[[Page 25657]]
standards and applicable safety criteria limits are met. The
proposed amendment would increase the time during which the
equipment hatch and personnel air locks could be open during core
alterations and movement of irradiated fuel. The proposed amendment
does not involve changes in the operations of these containment
penetrations. Having these penetrations open does not create the
possibility of a new accident.
Therefore, operation of the Point Beach Nuclear Plant in
accordance with the proposed amendments will not create the
possibility of a new or different type of accident from any accident
previously evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant reduction
in a margin of safety.
The assumptions and input used in the analysis are conservative
as noted below. The design basis FHA has been defined to identify
conservative conditions. The source term and radioactivity releases
have been calculated pursuant to RG 1.183, Appendix B and with
conservative assumptions concerning prior reactor operations. The
control room atmospheric dispersion factor has been calculated with
conservative assumptions associated with the release. The
conservative assumptions and input noted above ensure that the
radiation doses cited in the amendment request are the upper bound
to radiological consequences of a FHA either in containment or in
the spent fuel pool. The analysis shows that there is a significant
margin between the TEDE radiation doses calculated for the
postulated FHA using the AST and acceptance limits of 10 CFR 50.67
and RG 1.183. The proposed changes will not degrade the plant
protective boundaries, will not cause a release of fission products
to the public, and will not degrade the performance of any
Structures, Systems, and Components important to safety. Therefore,
there is no significant reduction in the margin of safety as a
result of the proposed changes.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: L. Raghavan.
Southern Nuclear Operating Company, Inc, Docket No. 50-364, Joseph M.
Farley Nuclear Plant, Unit 2, Houston County, Alabama
Date of amendment request: February 11, 2003.
Description of amendment request: The proposed amendments would
allow a 40-month inspection interval for Farley, Unit 2 after the
completion of the first post-replacement in-service inspection, rather
than the completion of two consecutive inspections resulting in a
classification of C-1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed one-time change revises the steam generator (SG)
inspection interval requirements in TS [technical specification]
5.5.9.3, ``Inspection Frequencies,'' for the FNP [Farley Nuclear
Plant] Unit 2 Spring 2004 refueling outage, to allow a 40[-]month
inspection frequency after one inspection, rather than after two
consecutive inspections with results that are within the C-1
category. C-1 category is defined as ``less than 5% of the total
tubes inspected are degraded tubes and none of the inspected tubes
are defective.''
The proposed one-time extension of the FNP Unit 2 SG tube
inservice inspection interval does not involve changing any
structure, system, or component, or affect reactor operations. It is
not an initiator of an accident and does not change any existing
safety analysis previously analyzed in the FNP's Final Safety
Analysis Report (FSAR). As such, the proposed change does not
involve a significant increase in the probability of an accident
previously evaluated.
Since the proposed change does not alter the plant design, there
is no direct increase in SG leakage. Industry experience indicates
that the probability of increased SG tube degradation would not go
undetected. Additionally, steps described below will further
minimize the risk associated with this extension. For example, the
scope of inspections performed during the last FNP Unit 2 refueling
outage (i.e., the first refueling outage following SG replacement)
exceeded the TS requirements for the first two refueling outages
after SG replacement. That is, more tubes were inspected than were
required by TS. Currently, FNP Unit 2 does not have a SG damage
mechanism, and will meet the current industry examination guidelines
without performing SG inspections during the next refueling outage.
Additionally, as part of the FNP SG Program, both a Condition
Monitoring Assessment and an Operational Assessment are performed
after each inspection and compared to the Nuclear Energy Institute
(NEI) 97-06, ``Steam Generator Program Guidelines,'' performance
criteria. The results of the Condition Monitoring Assessment
demonstrated that all performance criteria were met during the FNP
Unit 2 Fall 2002 refueling outage, and the results of the
Operational Assessment show that all performance criteria will be
met over the proposed operating period. Considering these actions,
along with the improved SG design and reliability of Westinghouse
replacement SGs, extending the SG tube inspection frequency does not
involve a significant increase in 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?
The proposed change revises the SG inspection frequency
requirements in TS 5.5.9.3.a for the FNP Unit 2 Spring 2004
refueling outage, to allow a 40[-]month inspection interval after
one inspection, rather than after two consecutive inspections, with
inspection results within the C-1 category.
The proposed change will not alter any plant design basis or
postulated accident resulting from potential SG tube degradation.
The scope of inspections performed during the last FNP Unit 2
refueling outage (i.e., the first refueling outage following SG
replacement) significantly exceeded the TS requirements for the
scope of the first two refueling outages after SG replacement.
Primary-to-secondary leakage that may be experienced during all
plant conditions is expected to remain within current accident
analysis assumptions. The proposed change does not affect the design
of the SGs, the method of SG operation, or reactor coolant chemistry
controls. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. The
proposed change involves a one-time extension to the SG tube
inservice inspection frequency and therefore will not give rise to
new failure modes. In addition, the proposed change does not impact
any other plant systems or components.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The SG tubes are an integral part of the Reactor Coolant System
(RCS) pressure boundary that are relied upon to maintain the RCS
pressure and inventory. The SG tubes isolate the radioactive fission
products in the reactor coolant from the secondary system. The
safety function of the SGs is maintained by ensuring the integrity
of the SG tubes. In addition, the SG tubes comprise the heat
transfer surface between the primary and secondary systems such that
residual heat can be removed from the primary system.
SG tube integrity is a function of the design, environment, and
current physical condition. Extending the SG tube inservice
inspection frequency by one operating cycle will not alter the
function or design of the SGs. SG inspections conducted during the
first refueling outage following SG replacement demonstrated that
the SGs do not have an active damage mechanism, and the scope of
those inspections significantly exceeded the scope required by the
TS. These inspection results were comparable to similar inspection
results for second generation alloy 690 models of replacement SGs
installed at other plants, and subsequent inspections at those
plants yielded results that support this extension request. The
improved design of the replacement SGs also provides reasonable
assurance that significant tube degradation is not likely to occur
over the proposed operating period.
The NRC staff has reviewed the licensee's analysis and, based on
this
[[Page 25658]]
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Section Chief: John A. Nakoski.
Southern Nuclear Operating Company, Inc, Docket No. 50-364, Joseph M.
Farley Nuclear Plant, Unit 2, Houston County, Alabama
Date of amendment request: March 31, 2003.
Description of amendment request: The proposed amendments would
modify Surveillance Requirement (SR) 3.4.11.1, for Farley, Unit 2 only
by the addition of the following note that states, ``Not required to be
performed for Unit 2 for the remainder of operating cycle 16 for
Q2B31MOV800B.'' In addition, a temporary Technical Specification SR
3.4.11.4 is added to provide compensatory action for this block valve
while SR 3.4.11.1 is suspended. Further, this SR requires that power to
the Farley, Unit 2 Power Operated Relief Valve Q2B31MOV800B be checked
at least every 24 hours for the remainder of operating cycle 16.
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 changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change to Surveillance Requirement (SR) 3.4.11.1
suspends the requirement to cycle test the Unit Two pressurizer
power operated relief valve (PORV) block valve Q2B31MOV8000B for the
remainder of operating cycle 16. This change will eliminate the
remaining scheduled cycle tests for the PORV block valve during
operating cycle 16. SR 3.4.11.4 is added to provide compensatory
measures for verifying power available to the block valve at least
every 24 hours. At the end of cycle 16, the proposed changes will no
longer be in effect. Suspension of the cycle tests for the PORV
block valve Q2B31MOV8000B may result in a small decrease in
assurance that the block valve would cycle if required to isolate a
stuck open PORV. However, experience with these valves has shown
them to be very reliable and suspension of the remaining tests will
not appreciably reduce reliability of the valve. There is no
relationship between packing leakage on the PORV block valve and a
postulated stuck open PORV. The proposed compensatory measure of
verifying block valve power available on a 24 hour basis adds
additional assurance that the block valve will close if demanded.
Therefore, the probability of a previously evaluated accident
remains acceptable is not significantly increased.
The proposed changes do not affect the consequences of a
previously analyzed accident since the magnitude and duration of
analyzed events are not impacted by this change. The dose
consequences of the proposed change are bounded by LOCA [loss-of-
coolant accident] analyses. Therefore, the consequences of a
previously evaluated accident are unchanged.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes involve no change to the physical plant.
They allow for suspension of the PORV block valve Q2B31MOV8000B
cycle tests for a limited time and provide for compensatory action
to verify power to the PORV block valve. This valve provides an
isolation function for a postulated stuck open or leaking
pressurizer PORV. This condition is an analyzed event since it is
bounded by the FNP [Farley Nuclear Plant] LOCA analyses. In addition
to the isolation function, the block valve is required to remain
open to allow the associated PORV to function automatically to
control reactor coolant system (RCS) pressure. These changes do not
impact the open function of the block valve since the normal
position is open.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The physical plant is unaffected by these changes. The proposed
changes do not impact accident offsite dose, containment pressure or
temperature, emergency core cooling system (ECCS) or reactor
protection system (RPS) settings or any other parameter that could
affect a margin of safety. The elimination of cycle testing of the
PORV block valve Q2B31MOV8000B for the remainder of the Unit Two
operating cycle and the addition of the proposed compensatory action
that enhances assurance of valve operation are somewhat offsetting.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Section Chief: John A. Nakoski.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of amendment request: February 26, 2003.
Description of amendment request: The proposed amendments would
revise Technical Specifications Section 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 2002 for Unit 1 and March 1995 for Unit 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. The proposed Technical Specifications change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
The proposed revision to Technical Specifications 5.5.17,
``Containment Leakage Rate Testing Program,'' involves a one-time
extension to the current interval for Type A containment leak
testing. The current test interval of ten (10) years would be
extended on a one-time basis to no longer than fifteen (15) years
from the last Type A test. The proposed Technical Specifications
change does not involve a physical change to the plant or a change
in the manner in which the plant is operated or controlled. The
reactor containment is designed to provide an essentially leak tight
barrier against the uncontrolled release of radioactivity to the
environment for postulated accidents. As such, the reactor
containment itself and the testing requirements invoked to
periodically demonstrate the integrity of the reactor containment
exist to ensure the plant's ability to mitigate the consequences of
an accident, and do not involve the prevention or identification of
any precursors of an accident.
The proposed change involves only the extension of the interval
between Type A containment leakage tests. Type B and C containment
leakage tests will continue to be performed at the frequency
currently required by plant Technical Specifications. Industry
experience has shown, as documented in NUREG-1493 [``Performance-
Based Containment Leak-Test Program''], that Type B and C
containment leakage tests have identified a very large percentage of
containment leakage paths and that the percentage of containment
leakage paths that are detected only by Type A testing is very
small. VEGP [Vogtle Electric Generating Plant] test history supports
this conclusion. NUREG-1493 concluded, in part, that reducing the
frequency of Type A containment leak tests to once per twenty (20)
years leads to an imperceptible increase in risk. The integrity of
the reactor containment is subject to two types of failure mechanism
which can be categorized as (1) activity based and (2) time based.
Activity based failure mechanisms are defined as degradation due to
system and/or component modifications or maintenance. Local leak
rate test requirements and administrative controls such as design
change control and procedural
[[Page 25659]]
requirements for system restoration ensure that containment
integrity is not degraded by plant modifications or maintenance
activities. The design and construction requirements of the reactor
containment itself combined with the containment inspections
performed in accordance with ASME Section XI, the Maintenance Rule,
and the containment coatings program serve to provide a high degree
of assurance that the containment will not degrade in a manner that
is detectable only by Type A testing.
2. The proposed Technical Specifications change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
The proposed revision to Technical Specifications involves a
one-time extension to the current interval for Type A containment
leak testing. The reactor containment and the testing requirements
invoked to periodically demonstrate the integrity of the reactor
containment exist to ensure the plant's ability to mitigate the
consequences of an accident and do not involve the prevention or
identification of any precursors of an accident. The proposed
Technical Specifications change does not involve a physical change
to the plant or the manner in which the plant is operated or
controlled.
3. The proposed Technical Specifications change does not involve
a significant reduction in a margin of safety.
The proposed revision to Technical Specifications involves a
one-time extension to the current interval for Type A containment
leak testing. The proposed Technical Specifications change does not
involve a physical change to the plant or a change in the manner in
which the plant is operated or controlled. The specific requirements
and conditions of the Containment Leakage Rate Testing Program, as
defined in Technical Specifications, exist to ensure that the degree
of reactor containment structural integrity and leak tightness that
is considered in the plant safety analysis is maintained. The
overall containment leakage rate limit specified by Technical
Specifications is maintained. The proposed change involves only the
extension of the interval between Type A containment leakage tests.
Type B and C containment leakage tests will continue to be performed
at the frequency currently required by plant Technical
Specifications.
VEGP and industry experience strongly support the conclusion
that Type B and C testing detects a large percentage of containment
leakage paths and that the percentage of containment leakage paths
that are detected only by Type A testing is small. The containment
inspections performed in accordance with ASME Section XI, the
Maintenance Rule, and the containment coatings program serve to
provide a high degree of assurance that the containment will not
degrade in a manner that is detectable only by Type A testing.
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. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Section Chief: John A. Nakoski.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendments: April 8, 2003 as supplemented
April 22, 2003.
Brief description of amendments: To revise, for one time only, a
portion of Surveillance Requirement 3.5.2.3 of the Technical
Specifications for the emergency core cooling system (ECCS). The
revision will extend, until the refueling outage in the fall of 2003,
the verification that the ECCS safety injection hot leg injection lines
are full of water.
Date of publication of individual notice in the Federal Register:
April 16, 2003 (68 FR 18712).
Expiration date of individual notice: May 1, 2003
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].
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: September 26, 2002.
Brief description of amendment: The amendment revises Technical
Specification 3.9.1, ``Refueling Equipment Interlocks,'' to allow in-
vessel fuel movement to continue if the refueling interlocks become
inoperable. Specifically, the amendment adds Required Action A.2.1 to
immediately block control rod withdrawal and Required Action A.2.2 to
perform a verification that all of the control rods are fully inserted.
Date of issuance: April 28, 2003.
[[Page 25660]]
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 154.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 26, 2002 (67
FR 70764).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 28, 2003.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: December 4, 2002.
Brief description of amendments: The amendments revised the
Technical Specification 3.7.6 to require a minimum combined inventory
of 155,000 gallons and remove the Condensate Storage Tank as a source
of the combined inventory.
Date of Issuance: April 30, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 330, 330 & 331.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 31, 2003 (68 FR
2801).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 30, 2003.
No significant hazards consideration comments received: No
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: December 6, 2002.
Brief description of amendment: The amendment increased the
surveillance interval of the local power range monitor calibrations
from 1000 megawatt-days/ton to 200 megawatt-days/ton.
Date of issuance: May 1, 2003.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 277.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5674).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 1, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: August 16, 2002, as supplemented
on March 26, April 16, and April 19, 2003.
Brief description of amendment: The amendment modified Technical
Specification (TS) 3/4.10.A, ``Refueling Interlocks,'' and TS 3/4.10.D,
``Multiple Control Rod Removal,'' to provide an alternative required
action if the refueling interlocks became inoperable during fuel
movements in the reactor vessel. The amendment allowed fuel movements
to continue in the reactor vessel should the refueling equipment
interlocks become inoperable.
Date of issuance: April 21, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 199.
Facility Operating License No. DPR-35: Amendment revised the TSs.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75872).
The March 26, April 16, and April 19, 2003, letters provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated April 21, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: January 23, 2003, as
supplemented February 24, and April 17, 2003.
Brief description of amendment: This amendment modifies the Pilgrim
Nuclear Power Station Technical Specification (TS) requirements for the
Emergency Core Cooling System (ECCS) during shutdown conditions. The
amendment changes the Core Spray and Low Pressure Coolant Injection
System's TS requirements to be applicable during the Run, Startup, and
Hot Shutdown Modes. The amendment also modifies the High Drywell
Pressure Instrumentation TSs to require the instrumentation to be
Operable during the Run, Startup and Hot Shutdown Modes. Unnecessary TS
requirements are removed based on the plant's operating Mode. Other
changes are administrative in nature.
Date of issuance: April 22, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 200.
Facility Operating License No. DPR-35: Amendment revised the TSs.
Date of initial notice in Federal Register: March 18, 2003 (68 FR
12952).
The supplements dated February 24, and April 17, 2003, provided
additional information that clarified the application, and did not
expand the scope of the application or 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 22, 2003.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: September 18, 2002.
Brief description of amendment: The amendment revises several
Technical Specifications Limiting Conditions for Operations and
Administrative sections to correct or clarify certain requirements and
information.
Date of issuance: April 23, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No: 157.
Facility Operating License No. NPF-29: The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75871).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 23, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237, Dresden Nuclear
Power Station, Unit 2, Grundy County, Illinois
Date of application for amendment: January 31, 2003, as
supplemented March 7, 2003.
Brief description of amendment: The amendment revises the safety
limit
[[Page 25661]]
minimum critical power ratio for Unit 2 for two loop operation and for
single loop operation.
Date of issuance: April 22, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 199.
Facility Operating License No. DPR-19: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 4, 2003 (68 FR
10279). The supplement dated March 7, 2003, 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 22, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: April 10, 2002, as supplemented
March 10, 2003.
Brief description of amendments: The amendments revised the
Technical Specifications to relocate emergency diesel generator
maintenance inspection requirements from Section 4.8.1.1.2.e.1 to the
Technical Requirements Manual.
Date of issuance: April 18, 2003.
Effective date: As of the date of issuance and shall include the
relocation of the emergency diesel generator maintenance requirements
of Technical Specification 4.8.1.1.2.e.1 to the Technical Requirements
Manual within 30 days.
Amendment Nos.: 165 and 128.
Facility Operating License Nos. NPF-39 and NPF-85: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 28, 2002 (67 FR
36926). The supplement dated March 10, 2003, 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 amendments is contained
in a Safety Evaluation dated April 18, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: November 27, 2002.
Brief description of amendments: These amendments deleted TS
6.8.4.c, ``Post-accident Sampling,'' and thereby eliminated the
requirements to have and maintain the post accident sampling system for
Limerick Generating Station, Units 1 and 2. The amendments also
addressed related changes to TS 6.8.4.a, ``Primary Coolant Sources
Outside Containment.''
Date of issuance: April 25, 2003.
Effective date: As of date of issuance and shall be implemented
within 180 days.
Amendment Nos.: 166 and 129.
Facility Operating License Nos. NPF-39 and NPF-85. 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 April 25, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: June 4, 2002.
Brief description of amendment: This amendment revises the pressure
temperature limits for 22- and 32-effective full power years for Perry
Nuclear Power Plant. The June 4, 2002, application also contained a
request for exemption from applying Appendix G of the 1995 American
Society of Mechanical Engineers Boiler and Pressure Vessel Code and
approval for using Code Case N-640, which permits the use of the plain
strain fracture toughness (KIc) curve instead of the crack arrest
fracture toughness (KIa) curve for reactor pressure vessel materials in
determining the P-T limits.
Date of issuance: April 29, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 127.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75878).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 29, 2003.
No significant hazards consideration comments received: No.
Florida Power and Light Company, Docket No. 50-389, St. Lucie Plant,
Unit No. 2, St. Lucie County, Florida
Date of amendment request: October 15, 2002, as supplemented
February 28, 2003.
Description of amendment request: The amendment modifies the
reactor coolant system flow rate from 363,000 gallons per minute (gpm)
to 355,000 gpm in Technical Specifications (TSs) Table 3.3-2 and in a
footnote for Table 2.2-1.
Date of Issuance: April 18, 2003.
Effective Date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 131.
Facility Operating License No. NPF-16: Amendment revised the TSs.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68737).
The February 28, 2003, supplement did not affect the original
proposed no significant hazards determination, or expand the scope of
the request as noticed in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 18, 2003.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: July 23, 2002.
Brief description of amendments: The amendments revise certain 18-
month surveillance requirements by eliminating the condition that
testing be conducted ``during shutdown,'' or ``during the COLD SHUTDOWN
or REFUELING MODE'' (i.e., shutdown conditions).
Date of issuance: April 22, 2003.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 275 and 257.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 17, 2002 (67
FR 58647).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 22, 2003.
No significant hazards consideration comments received: No.
[[Page 25662]]
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: January 14, 2003.
Brief description of amendments: The amendments modify Technical
Specification (TS) 3.7.5.1 to add an exception to Limiting Condition
for Operation 3.0.4 for the control room emergency ventilation system
(CREVS). This exception allows movement of irradiated fuel assemblies
to begin while one of the two CREVS pressurization trains is
inoperable, provided the appropriate TS action requirements are
implemented. The amendments are consistent with the standard TSs for
Westinghouse plants (NUREG 1431, Revision 2, ``Standard Technical
Specifications, Westinghouse Plants,'' dated April 30, 2001).
Date of issuance: April 25, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 276 and 258.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 4, 2003 (68 FR
10280).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 25, 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: October 26, 2001, as
supplemented by letters dated June 7 and November 22, 2002.
Brief description of amendment: The amendment revised Section 6.0,
``Administrative Controls,'' of the Technical Specifications (TSs) to
clarify and relocate existing requirements, make wording improvements,
and make the TSs consistent with the Unit 2 TSs. The revised Section
6.0 is consistent with the ``Standard Technical Specifications for
General Electric plants, BWR [Boiling Water Reactor]/4'' (NUREG-1433,
Revision 2).
Date of issuance: April 23, 2003.
Effective date: April 23, 2003, to be implemented within 90 days of
issuance.
Amendment No.: 181.
Facility Operating License No. DPR-63: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 8, 2002 (67 FR
928).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 23, 2003.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: July 26, 2002, as supplemented
February 27, March 14, March 19, March 21 (2 letters), and April 3,
2003.
Brief description of amendment: The amendment revises technical
specifications for use of Westinghouse 422 VANTAGE + nuclear fuel with
PERFORMANCE + features.
Date of issuance: April 4, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 167.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 3, 2002 (67
FR 56322).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 4, 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: September 19, 2002, as
supplemented February 28, 2003.
Brief description of amendment: The amendment relocates TS
Surveillance Requirement (SR) 4.6.B.2, ``Reactor Vessel Temperature and
Pressure,'' and the associated TS Bases to Section 4.2 of the Updated
Safety Analysis Report. It also implements the Boiling Water Reactor
Vessel and Internals Project reactor pressure vessel integrated
surveillance program at Monticello and demonstrates compliance with the
requirements of Title 10 of the Code of Federal Regulations, Part 50,
Appendix H, ``Reactor Vessel Material Surveillance Program
Requirements.''
Date of issuance: April 22, 2003
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 135.
Facility Operating License No. DPR-22: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 29, 2002 (67 FR
66012).
The supplement of February 28, 2003, 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 April 22, 2003.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: October 8, 2002, and its supplements
dated December 3, 2002, and March 4, 2003.
Brief description of amendment: The amendment modifies Technical
Specification 2.3.a, ``Emergency Core Cooling System,'' to extend the
allowed outage time for a single low pressure safety injection pump
from the existing 24 hours to 7 days. In addition, the word ``pump''
has been replaced with the word ``train.''
Date of issuance: April 29, 2003
Effective date: April 29, 2003, and shall be implemented within 60
days from the date of issuance.
Amendment No.: 217.
Facility Operating License No. DPR-40: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68740).
The supplemental letters dated December 3, 2002, and March 4, 2003,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed or revise the
proposed technical specification changes 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 29, 2003.
No significant hazards consideration comments received: No.
Saxton Nuclear Experimental Corporation (SNEC) and GPU Nuclear, Inc.,
Docket No. 50-146 Saxton Nuclear Experimental Facility (SNEF), Bedford
County, Pennsylvania
Date of application for amendment: February 2, 2000, as
supplemented on
[[Page 25663]]
June 23, August 11, September 18 and December 4, 2000; January 30,
February 14, March 15 and 19, June 20, July 2 and September 4, 2001;
and January 11 and 24, February 4, May 22 and 28, July 11, August 20,
September 17, 23, 24, and 26, October 10, and December 16, 2002.
Brief description of amendment: The amendment revises Amended
Facility License No. DPR-4 for the SNEF to annotate approval of the
SNEF License Termination Plan.
Date of issuance: March 28, 2003.
Effective date: Date of issuance to be implemented no later than 30
days from the date of issuance.
Amendment No.: 18.
Amended Facility License No. DPR-4: Amendment added a new license
condition to require the licensees to implement and maintain in effect
all provisions of the approved SNEF License Termination Plan.
Date of initial notice in Federal Register: November 29, 2000.
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.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendments request: November 5, 2001, as supplemented by
letters dated October 23, 2002, and January 15, 2003.
Brief description of amendments: The proposed amendments convert
the current Technical Specification (TS) Section 6.0 of the STP, Units
1 and 2, TS to the Improved Technical Specifications based on NUREG-
1431, ``Standard Technical Specification for Westinghouse Plants.''
Date of issuance: April 24, 2003.
Effective date: As of its date of issuance and shall be implemented
within 6 months from the date of issuance.
Amendment Nos.: Unit 1-151; Unit 2-139.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 5, 2002 (67 FR
5335).
The October 23, 2002, and January 15, 2003, supplemental letters
provided clarifying information that was within the scope of the
original Federal Register notice (67 FR 5335) and did not change the
initial no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 24, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant,
Unit 1, Hamilton County, Tennessee
Date of application for amendment: February 28, 2003.
Brief description of amendment: The amendment approves the use of
an alternate methodology using a through-bolted connection frame to
restore the steam generator (SG) compartment roof after replacement of
the SGs, and a revision of the Updated Safety Analysis Report (UFSAR)
to reflect the approval of the methodology.
Date of issuance: April 25, 2003.
Effective date: As of the date of issuance, to be incorporated into
the UFSAR at the time of its next update.
Amendment No.: 184.
Facility Operating License No. DPR-77: Amendment revises the UFSAR.
Date of initial notice in Federal Register: March 14, 2003 (68 FR
12382).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 25, 2003.
No significant hazards consideration comments received: No
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant
(SQN), Unit 1, Hamilton County, Tennessee
Date of application for amendment: February 28, 2003.
Description of amendment: The amendment approves a revision of the
SQN Updated Final Safety Analysis (UFSAR) to include a change to the
methodology for connecting reinforcing steel bars during restoration of
the Unit 1 concrete shield building dome as part of the steam generator
replacement project. This modification to the shield building concrete
dome is necessary to support removal of the original steam generators
and installation of the replacement steam generators.
Date of issuance: April 24, 2003.
Effective date: As of the date of issuance to be incorporated into
the UFSAR at the time of its next update.
Amendment No.: 283.
Facility Operating License No. DPR-77: Amendment revises the
Operating License.
Date of initial notice in Federal Register: March 17, 2003 (68 FR
12718).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 24, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: April 8, 2003, as supplemented
April 22, 2003.
Brief description of amendment: The amendment revises, for one time
only, a portion of Surveillance Requirement (SR) 3.5.2.3 of the Watts
Bar Technical Specifications for the emergency core cooling system
(ECCS). The revision extends, until the refueling outage in the fall of
2003, the verification that the ECCS safety injection hot leg injection
lines are full of water. SR 3.5.2.3 currently requires a verification
frequency of 31 days.
Date of issuance: May 1, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 43.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes. (68 FR 18712 dated April 16, 2001). That notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by May 16, 2003, but indicated that if the Commission makes a
final no significant hazards consideration determination, any such
hearing would take place after issuance of the amendment. The April 22,
2003, letter provided clarifying information that did not change the
initial proposed no significant hazards consideration determination or
expand the scope of the original request.
The Commission's related evaluation of the amendments, finding of
exigent circumstances, and final no significant hazards consideration
determination are contained in a Safety Evaluation dated May 1, 2003.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: October 3, 2002.
Brief description of amendment: The amendment revises Limiting
Condition for Operation 3.1.8, ``Physics Tests Exceptions--Mode 2,'' to
reduce the required number of channels from four
[[Page 25664]]
to three channels for certain functions in Table 3.3.1-1, ``Reactor
Trip System Instrumentation.''
Date of issuance: April 21, 2003.
Effective date: April 21, 2003, and shall be implemented within 60
days of the date of issuance.
Amendment No.: 154.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 26, 2002 (67
FR 70771).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 21, 2003.
No significant hazards consideration comments received. No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: October 1, 2002.
Brief description of amendment: The amendment revises Limiting
Condition for Operation 3.1.8, ``Physics Tests Exceptions--Mode 2,'' to
reduce the required number of channels from four to three channels for
certain functions in Table 3.3.1-1, ``Reactor Trip System
Instrumentation.''
Date of issuance: April 21, 2003.
Effective date: April 21, 2003, and shall be implemented within 90
days of the date of issuance.
Amendment No.: 151.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68746).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 21, 2003.
No significant hazards consideration comments received: No.
Yankee Atomic Electric Co., Docket No. 50-29, Yankee Nuclear Power
Station (YNPS) Franklin County, Massachusetts
Brief description of amendment: The amendment revises the YNPS
License and Technical Specifications to delete operational and
administrative requirements that would no longer be required once the
spent nuclear fuel has been transferred from the spent fuel pool to the
Independent Spent Fuel Storage Installation.
Date of issuance: April 17, 2003.
Effective date: April 17, 2003.
Amendment No.: 157.
Facility Operating License No. DPR-3. Amendment revises the License
and Technical Specifications.
Date of initial notice in Federal Register: February 18, 2003 (68
FR 7823).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 17, 2003.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 5th day of May 2003.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 03-11697 Filed 5-12-03; 8:45 am]
BILLING CODE 7590-01-P