[Federal Register Volume 66, Number 182 (Wednesday, September 19, 2001)]
[Notices]
[Pages 48283-48295]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-23209]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
(Note: The publication date for this notice will change from
every other Wednesday to every other Tuesday, effective January 8,
2002. The notice will contain the same information and will continue
to be published biweekly.)
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 August 27, 2001 through September 7, 2001.
The last biweekly notice was published on September 5, 2001 (66 FR
46473).
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)
[[Page 48284]]
involve a significant reduction in a margin of safety. The basis for
this proposed determination for each amendment request is shown below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC Public Document
Room, located at One White Flint North, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. The filing of requests for a hearing
and petitions for leave to intervene is discussed below.
By October 19, 2001, 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 Public Document Room, located at One
White Flint North, 11555 Rockville Pike (first floor), Rockville,
Maryland 20852. Publicly available records will be accessible and
electronically from the ADAMS Public Library component on the NRC Web
site, http://www.nrc.gov (the Electronic Reading Room). 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 Branch, or may be delivered to the Commission's Public
Document Room, located at One White Flint North, 11555 Rockville Pike
(first floor), Rockville, Maryland 20852, by the above date. A copy of
the petition should also be sent to the Office of the General Counsel,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and 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
[[Page 48285]]
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 Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Assess and Management Systems (ADAMS) Public Electronic
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/NRC/ADAMS/index.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, 304-
415-4737 or by email to [email protected].
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: July 23, 2001.
Description of amendment request: As a follow-up response to a
commitment identified in the Nuclear Regulatory Commission (NRC) staff
letter dated December 22, 2000, ``Completion of Licensing Action for
Generic Letter (GL) 96-06, Assurance of Equipment Operability and
Containment Integrity During Design-Basis Accident Conditions,''
Entergy Operations Inc., (Entergy, the licensee) has proposed to revise
their Waterford Steam Electric Station, Unit 3 (Waterford 3) Final
Safety Analysis Report (FSAR) to resolve the ten containment
penetrations susceptible to thermally induces overpressurization
through an evaluation, detailed analysis, or installation of physical
modifications prior to startup from the spring 2002 refueling outage.
Entergy determined a change to Waterford 3's license basis, through
procedural controls, risk analysis, and engineering analysis, for seven
penetrations, as discussed in this license basis change request.
Permanent resolution to the GL 96-06 issues for the remaining three
penetrations could be satisfied through the installation of physical
modifications.
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. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
The proposed FSAR change reflects the use of administrative
procedural controls to ensure these seven containment penetrations
(two 4-inch diameter Steam Generator Blowdown penetrations and five-
\1/2\ inch diameter Process Sampling penetrations) contain fluid at
temperatures representative of Reactor Coolant, and the very low
probability for overpressurization failure of containment
penetrations during Mode 4 plant operation as a permanent solution
to the GL 96-06 issue. The engineering analysis determined these
seven containment penetrations met the acceptance criteria for
allowed stresses contained in ASME [American Society of Mechanical
Engineers] Section III Code, [Boiler and Pressure Vessel Code]
Appendix F 1995. The result of the risk analysis is such that the
very small change in LERF (Large Early Release Frequency], on the
order of 1 x 10-9 per reactor year, remained
well below the 1 x 10-7 LERF
guideline for a small change given in Regulatory Guide 1.174. The
negligible reduction in LERF that would be achieved by adding
thermal relief valve overpressure protection is not risk significant
and is too small to justify the addition of the relief valves.
With respect to the probability or the consequences of an
accident previously evaluated in the FSAR, the proposed deviation to
the existing ASME Section III Code, Class 2 design provisions and
operating requirements for the seven containment penetrations would
not significantly increase the probability of an accident since the
administrative procedural controls are being provided to: (1)
minimize penetration heat-up and over-pressurization during a small
window of vulnerability, approximately 1% per year of Mode 4 plant
operation; and (2) minimize process fluid cooldown during normal
plant operation by closing the containment isolation valves for the
five sample penetrations when process fluid samples are obtained and
the laboratory sample valves downstream of the CIV [containment
isolation valves] are closed or flow through the penetration is
stopped. Also the results of engineering analyses showed that the
containment penetrations may exceed ASME Section III, Subsection NC
3500 Code required yield stresses and experience plastic
deformation, but would not catastrophically fail; therefore, the
penetrations would retain their ability to perform their safety
function and maintain containment integrity.
On this basis, the proposed changes are not considered to
constitute a significant increase in the probability or consequences
of an accident due to:
Administrative controls to minimize penetration heat-up
and over-pressurization during the small window of vulnerability
The seven containment penetrations retaining their
ability to perform their safety function and maintaining containment
integrity in accordance with engineering analyses performed that met
acceptance criteria for allowed stresses contained in ASME Section
III Code, Appendix F 1995, and
The low risk significance of overpressurization failure
of the seven containment penetrations during a DBA [Design Basis
Accident] while the plant is in Mode 4.
The proposed changes will not significantly affect the results
of any accident previously evaluated. The accident mitigation
features of the plant are not significantly affected by these
proposed changes. The proposed changes do not add or modify any
existing equipment.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
The change proposes a deviation to the existing ASME, Section
III, Class 2 license basis requirements for portions of the Steam
Generator Blowdown System, Primary Sampling System, and Secondary
Sampling System that penetrate the containment as a permanent
solution to the GL 96-06 issues. This change involves recognition of
the acceptability of administrative procedural controls to minimize
penetration heat-up and over-pressurization during the small window
of vulnerability, approximately 1% per year for Mode 4 plant
operation. Added assurance is provided through the engineering
analysis performed on these penetrations that determined allowable
stresses did not exceed the ASME Section III Code, Appendix F 1995
pipe stress values. Therefore, the change would not contribute to
the possibility of, or be the initiator for any new or different
kind of accident.
The proposed change does not alter the configuration of the
plant. There has been no physical change to plant systems,
structures, or components.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
The proposed change does not involve a significant reduction in
margin of safety. The existing licensing basis for Waterford 3, with
respect to the ASME Section III, Subsection NC-3621.2 provisions for
portions of the Steam Generator Blowdown System, Primary Sampling
System, and Secondary Sampling System that penetrate the
containment, is to ensure piping that has the potential to
experience pressurization due to trapped fluid expansion shall be
designed to withstand the increased pressure or have provisions for
relieving the excess pressure piping. With the acceptance of this
proposed deviation to the license basis, it will be recognized that
the seven containment penetrations have administrative procedural
controls to minimize penetration heat-up and over-pressurization
during the small window of vulnerability, approximately 1% per year
for Mode 4 plant operation. Added assurance is also provided through
the engineering analysis performed on these penetrations that
[[Page 48286]]
determined stresses did not exceed the ASME Section III Code,
Appendix F 1995 pipe stress values and predicted the penetration
piping would experience plastic deformation, but would not
catastrophically fail. Therefore, the penetrations would retain
their ability to perform their safety function and maintain
containment integrity. This deviation to license basis requirements
for these seven containment penetrations is not considered to
constitute a significant decrease in the margin of safety.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn
1400 L Street NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County,
Pennsylvania
Date of amendment request: August 13, 2001.
Description of amendment request: The proposed amendments delete
requirements from the Technical Specifications (TSs) to maintain a
Post-Accident Sampling System (PASS). Licensees were generally required
to implement PASS upgrades as described in NUREG-0737, ``Clarification
of TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power
Plants to Assess Plant and Environs Conditions During and Following an
Accident.'' Implementation of these upgrades was an outcome of the
lessons learned from the accident that occurred at TMI, Unit 2.
Requirements related to PASS were imposed by Order for many facilities
and were added to or included in the 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 NRC staff issued a notice of opportunity for comment in the
Federal Register on August 11, 2000 (65 FR 49271) on possible
amendments to eliminate PASS, 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 October 31, 2000 (65 FR 65018). The licensee affirmed the
applicability of the following NSHC determination in its application
dated August 13, 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated.
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post-accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from 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 radionuclides
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in 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.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Timothy G. Colburn, Acting.
[[Page 48287]]
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: August 22, 2001.
Description of amendment request: Florida Power and Light Company
(FPL) requests to amend Facility Operating Licenses DPR-67 for St.
Lucie Unit I and NPF-16 for St. Lucie Unit 2 by revising Technical
Specifications (TS) relating to positive reactivity additions while in
shutdown modes. The proposed changes clarify TS involving positive
reactivity additions to the shutdown reactor, and would allow small,
controlled, safe insertions of positive reactivity while in shutdown
modes. The proposed changes conform closely to an NRC approved generic
change for Standard Technical Specifications, known as TSTF-286 Rev. 2,
which revises most actions requiring ``Suspend operations involving
positive reactivity additions'' to allow minimum reactivity additions
due to temperature fluctuations or operations, which are necessary to
maintain fluid inventory within the required shutdown margin or
refueling boron concentration, as applicable.
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 facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed TS changes revise actions that either require
suspension of operations involving positive reactivity additions or
preclude reduction in boron concentration less than the reactor
coolant system (RCS). Reactivity excursions are analyzed events. The
proposed changes limit positive reactivity additions into the RCS
such that the required shutdown margin (SDM) or refueling boron
concentration continue to be met. Reactivity changes performed
during shutdown modes are currently governed by strict
administrative controls. Although the proposed changes will allow
procedural flexibility with regards to RCS temperature and boron
concentration, these operations will still be under administrative
control. The changes proposed by these amendments are within the
scope and assumptions of the existing analyses. Therefore, operation
of the facility in accordance with the proposed amendments would not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed TS revisions relate to positive reactivity
additions while in shutdown modes of operation. Reactivity
excursions are analyzed events. The operational flexibility allowed
in these proposed license amendments will be performed under strict
administrative controls in order to limit the potential for excess
positive reactivity addition. Although the existing procedural
controls will need modification, no new or different operational
failure modes would be introduced by these changes.
Additionally, implementation of these proposed changes do not
require any physical plant modifications, so no new or different
hardware related failure modes are introduced. The changes proposed
by these amendments are within the scope and assumptions of the
existing analyses. Therefore, operation of the facility in
accordance with the proposed amendments would not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed changes conform closely to the industry and NRC
approved TSTF-286, Rev. 2 and relate to small, controlled, safe
insertions of positive reactivity additions while in shutdown modes.
These changes revise actions that either require suspension of
operations involving positive reactivity additions, or prohibit RCS
boron concentration reduction. The proposed changes provide
operational flexibility while controlling positive reactivity
additions in order to preserve the required SDM or refueling boron
concentration. The proposed changes to provide for continued safe
reactor operations, while also limiting any potential for excess
positive reactivity addition. Therefore, operation of the facility
in accordance with the proposed amendments would not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Richard P. Correia.
Florida Power and Light Company, Docket No. 50-389, St. Lucie Plant,
Unit No. 2, St. Lucie County, Florida
Date of amendment request: June 22, 2001, as supplemented August
24, 2001.
Description of amendment request: The proposed amendment would
revise the St. Lucie Unit 2 Technical Specification (TS) 3.9.4,
Containment Penetrations. TS 3.9.4.a. requires that the containment
equipment door be closed during core alterations or movement of
irradiated fuel within containment. TS 3.9.4.b. requires a minimum of
one door in each airlock to be closed during core alterations or
movement of irradiated fuel within containment. The proposed change to
TS 3.9.4.a. would allow the containment equipment door to be open
during core alterations and movement of irradiated fuel in containment
provided: (a) The equipment door is capable of being closed with four
bolts within 30 minutes, (b) the plant is in MODE 6 with at least 23
feet of water above the reactor pressure vessel flange, and (c) a
designated crew is available at the equipment door to close the door.
The capability to close the containment equipment door includes the
requirements that the door is capable of being closed and that any
cables or hoses across the equipment door have quick-disconnects to
ensure the door is capable of being closed in a timely manner. The
proposed change to TS 3.9.4.b would allow both doors of each
containment airlock to be open during core alterations and movement of
irradiated fuel in containment provided: (a) At least one door of each
open containment airlock is capable of being closed, (b) the plant is
in MODE 6 with at least 23 feet of water above the reactor pressure
vessel flange, and (c) a designated individual is available outside
each open containment airlock to close the door. The capability to
close the containment airlock door includes the requirement that the
door is capable of being closed and that any cables or hoses across the
airlock door have quick-disconnects to ensure the door is capable of
being closed in a timely manner.
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 facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change to TS 3.9.4 would allow the containment
equipment door and both doors of each containment airlock to be open
during fuel movement or core alterations. Currently, the equipment
door is closed with four (4) bolts and a single door on each
containment airlock is closed during fuel movement or core
alterations to prevent the escape of radioactive material in the
[[Page 48288]]
event of an in-containment fuel handling accident. Neither the
containment equipment door nor either of the containment airlock
doors is an initiator of an accident. Whether the containment
equipment door or both doors of the containment air locks are open
or closed during fuel movement and core alterations has no affect on
the probability of any accident previously evaluated. Allowing the
containment equipment door and the containment airlock doors to be
open during fuel movement or core alterations does not significantly
increase the consequences from a fuel handling accident. The
calculated offsite doses are well within the limits of 10 CFR part
100. In addition, the calculated doses are larger than the expected
doses because the calculation does not incorporate the closing of
the containment equipment door or the containment airlock doors
after the containment is evacuated, which would be much less than
the two hours assumed in the analysis. The proposed change would
significantly reduce the dose to workers in containment in the event
of a fuel handling accident by reducing the time required to
evacuate the containment. The changes being proposed do not affect
assumptions contained in other plant safety analyses or the physical
design of the plant, nor do they affect other Technical
Specifications that preserve safety analysis assumptions. Therefore,
operation of the facility in accordance with the proposed amendments
would not involve a significant increase in the probability or
consequences of an accident previously analyzed.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed change to Technical Specification 3.9.4,
``Containment Building Penetrations,'' affects a previously
evaluated fuel handling accident. The new Fuel Handling Accident
Analysis assumes that all of the iodine and noble gases that become
airborne escape and reach the exclusion boundary and low population
zone with no credit taken for filtration, the containment building
barrier or for decay or deposition. Since the proposed change does
not involve the addition or modification of equipment nor does it
alter the design of plant systems and the revised analysis is
consistent with the Fuel Handling Accident Analysis, the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The margin of safety as defined by 10 CFR part 100 has not been
significantly reduced. The calculated dose is well within the limits
given in 10 CFR part 100 or NUREG 0800. The proposed change does not
alter the bases for assurance that safety-related activities are
performed correctly or the basis for any Technical Specification
that is related to the establishment of or maintenance of a safety
margin. Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Richard P. Correia.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: February 28, 2001.
Description of amendment request: The proposed amendment to the
Cooper Nuclear Station (CNS) Operating License DPR-46 would revise the
design basis accidents (DBA) radiological assessment methodology for
offsite and control room radiological doses, and the associated
supporting Technical Specifications (TS).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed revisions to the CNS DBA radiological assessment
methodology for offsite and control room doses, and the associated
supporting TS changes, do not involve initiators or precursors of
accidents previously evaluated. Furthermore, these changes do not
affect the design, function, or modes of operation of systems,
structures, or components within the facility. Therefore, the
proposed radiological assessment calculational methodology revisions
and TS changes do not involve a significant increase in the
probability of an accident previously evaluated in the Updated
Safety Analysis Report (USAR).
The proposed revisions to the CNS DBA radiological assessment
methodology for offsite and control room doses, and the associated
supporting TS changes, do not affect the design, function or modes
of operation of systems, structures or components in the facility.
The calculation revisions utilize conservatively lower accident
mitigation system filter efficiency assumptions and incorporate
plant specific accident mitigation system operating parameter and
design assumptions. Due to the changes in the calculational
methodology and assumptions, and an increase in the postulated
accident source term, the calculated radiological dose consequences
of each DBA have changed and in some cases increased. In each case,
however, the calculated radiological dose consequences are within
the exclusion area boundary (EAB) and low population zone (LPZ)
radiological dose acceptance criteria specified in 10 CFR part 100
and the control room dose acceptance criteria discussed in General
Design Criterion (GDC) 19 of 10 CFR part 50, Appendix A. Therefore,
the proposed revisions to the radiological assessment methodology,
and associated TS changes, do not involve a significant increase in
the consequences of an accident previously evaluated in the USAR.
2. Does not create the possibility for a new or different kind
of accident from any accident previously evaluated.
The proposed revisions to the CNS DBA radiological assessment
methodology for offsite and control room doses, and the associated
supporting TS changes, do not affect the design, function or mode of
operation of systems, structures or components in the facility such
that new equipment failure modes are created. No new or different
type of plant equipment is installed by the revised radiological
assessment calculational methodology or changes to the TS. Neither
the calculations nor the TS changes introduce changes to existing
design parameters governing normal plant operation or new plant
operating modes. No new types of accident initiators or precursors
are created by the proposed revisions. Therefore, the proposed
revisions to radiological assessment methodology and the proposed
changes to the TS do not create the possibility of a new or
different kind of accident previously evaluated in the USAR.
3. Does not create a significant reduction in the margin of
safety.
The proposed revisions to the CNS DBA radiological assessment
methodology for offsite and control room doses, and the associated
supporting TS changes, do not affect the design, function or mode of
operation of systems, structures or components in the facility.
These proposed TS changes are consistent with the criteria of 10 CFR
50.36(c)(2)(ii) for TS content.
The proposed revisions will not result in any challenges to
plant equipment, fuel integrity, or the reactor coolant system
pressure boundary. Due to the changes in the calculational
methodology and assumptions, and an increase in the postulated
accident source term, the calculated radiological dose consequences
of each design basis accident have changed and in some cases
increased. In each case, however, the calculated radiological dose
consequences are within the EAB and LPZ radiological dose acceptance
criteria specified in 10 CFR part 100 and the control room dose
acceptance criteria discussed in GDC 19 of 10 CFR part 50, Appendix
A. Therefore, the proposed revisions to the radiological assessment
methodology, and associated TS 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
[[Page 48289]]
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: April 12, 2001.
Description of amendment request: The proposed amendment would
change the Cooper Nuclear Station (CNS) Technical Specification (TS)
5.5.10.b.2 to replace the phrase, ``A change to the updated FSAR or
Bases that involves an unreviewed safety question as defined in 10 CFR
50.59'' with the phrase ``A change to the updated FSAR or Bases that
requires NRC approval pursuant to 10 CFR 50.59.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change deletes the reference to unreviewed safety
question as defined in 10 CFR 50.59. Deletion of the definition of
unreviewed safety question was approved by the NRC with the
revisions to 10 CFR 50.59. Consequently, the probability of an
accident previously evaluated is not significantly increased.
Changes to the TS Bases are still evaluated in accordance with 10
CFR 50.59. As a result, the consequences of any accident previously
evaluated are not significantly affected. Therefore, this change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
The proposed change will not reduce the margin of safety because
it has no direct effect on any safety analyses assumptions. Changes
to the TS Bases that result in meeting the criteria in revised 10
CFR 50.59 (c)(2) will still require NRC approval pursuant to 10 CFR
50.59. This change is administrative in nature as discussed by the
NRC in FR (Volume 64, Number 191, Pages 53582-53617) dated October
4, 1999, docketing the change to 10 CFR 50.59. 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. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: April, 12, 2001.
Description of amendment request: The amendment request would
modify the Cooper Nuclear Station (CNS) Technical Specifications
Surveillance Requirement (SR) 3.6.1.3.8 to relax the SR frequency by
allowing a representative sample of Excess Flow Check Valves (EFCVs) to
be tested every 18 months, such that each EFCV will be tested 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. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The current SR frequency requires each reactor instrumentation
line EFCV to be tested every 18 months. The EFCVs at CNS are
designed to close automatically in the event of a line break
downstream of the valve. This proposed change allows a reduced
number of EFCVs to be tested every 18 months. Industry operating
experience, documented in BWR [Boiling Water Reactor] Owners' Group
Topical Report NEDO-32977-A [``Excess Flow Check Valve Testing
Relaxation,'' dated June 2000], concludes that a change in
surveillance test frequency has a minimal impact on the reliability
for these valves. A failure of an EFCV to isolate cannot initiate
previously evaluated accidents. Furthermore, neither the EFCV
actuation test, nor the frequency of testing is considered an
initiator of any analyzed event. Therefore, there is no increase in
the probability of occurrence of an accident as a result of this
proposed change.
The consequences of a previously analyzed event are dependent on
the initial conditions assumed for the analysis, and the
availability and successful functioning of the equipment assumed to
operate in response to the analyzed event, and the setpoints at
which these actions are initiated. This change does not affect the
performance of any credited equipment. The installed restricting
orifice on each associated instrument line provides assurance that
any instrument line break will limit offsite doses to substantially
below 10 CFR part 100 values. Neither the EFCV actuation test, nor
the frequency of testing is an analysis assumption. Therefore, there
is no increase in the previously evaluated consequences of the
rupture of an instrument line and there is no potential increase in
the radiological consequences of an accident previously evaluated as
a result of this change.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
This proposed change allows a reduced number of EFCVs to be
tested each operating cycle. No other changes in requirements are
being proposed. Industry operating experience as documented in [BWR
Owners' Group Topical Report NEDO-32977-A] provides supporting
evidence that the reduced testing frequency 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 test frequency is
bounded by the previous evaluation of an instrument line pipe break.
This change will not physically alter the plant (no new or different
type of equipment will be installed). This change will not alter the
operation of process variables, structures, systems, or components
as described in the safety analysis. Thus, a new or different kind
of accident will not be created.
3. Does this change involve a significant reduction in a margin
of safety?
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. EFCV design, operation, and flow actuation criteria
remain unaffected by this change. Restricting orifices for each
associated instrument line remains available to mitigate an
instrument line break. The proposed change, which impacts the
frequency of testing EFCVs is acceptable because the tests continue
to require appropriate confirmation of the assumed function of the
system (and thereby assure continued operability), and has been
shown to reflect an acceptable frequency for detecting failures.
There is no detrimental impact on any other equipment design
parameter, and the plant will still be required to operate within
prescribed limits. Therefore, the change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
[[Page 48290]]
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: August 9, 2001.
Description of amendment request: The amendment would change the
Seabrook Station Technical Specifications (TSs) Index, TS 3/4.9.3
(``Decay Time''), TS 3/4.9.4 (``Containment Building Penetrations''),
and TS 3/4.9.9 (``Containment Purge And Exhaust Isolation System'').
The amendment would also change Bases 3/4.9.3, Bases 3/4.9.4, and Bases
3/4.9.9 for consistency with the proposed TS changes. These changes are
consistent with the improved Standard Technical Specifications (STS)
for Westinghouse plants.
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 changes to TS Index, TS 3/4.9.3, TS 3/4.9.4, and TS
3/4.9.9 do not adversely affect accident initiators or precursors
nor do they adversely alter the design assumptions, conditions, and
configuration of the facility or the manner in which the plant is
operated and maintained. In addition, the proposed changes do not
adversely affect the manner in which the plant responds in normal
operation, transient or accident conditions nor do they change any
of the procedures related to operation of the plant. Though a
portion of the proposed change to TS 3/4.9.4 appears to be a
relaxation to the current licensing basis, North Atlantic has
incorporated administrative conservatism into TS 3/4.9.4 to assure
the proposed changes, in conjunction with other TS required
surveillance testing, do not alter or prevent the ability of
structures, systems and components (SSCs), in particular the
Containment Purge and Exhaust Isolation System, to perform its
intended function to mitigate the consequences of an initiating
event within the acceptance limits assumed in the Updated Final
Safety Analysis Report (UFSAR).
The proposed changes do not adversely affect the source term,
containment isolation or radiological release assumptions used in
evaluating the radiological consequences of an accident previously
evaluated in the Seabrook Station UFSAR. Further, the proposed
changes do not increase the types and amounts of radioactive
effluent that may be released offsite, nor significantly increase
individual or cumulative occupational/public radiation exposures.
Therefore, it is concluded that these proposed revisions to TS
Index, TS 3/4.9.3, TS 3/4.9.4, and TS 3/4.9.9 do not involve a
significant increase in the probability or consequence of an
accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
This proposed changes to TS Index, TS 3/4.9.3, TS 3/4.9.4, and
TS 3/4.9.9 do not adversely affect the operation nor do they change
the design basis of any plant system or component during normal or
accident conditions. The proposed changes do not include any
physical changes to the plant. In addition, the proposed changes do
not adversely affect the function or operation of plant equipment or
introduce any new failure mechanisms such that the design basis is
adversely affected. The current licensing basis allows penetration
isolation by manual or automatic means. The plant equipment will
continue to respond per the design and analyses and there will not
be a malfunction of a new or different type introduced by the
proposed changes that creates the possibility of a new or different
kind of accident.
The proposed changes do not modify the facility nor do they
adversely affect the plant's response to normal, transient or
accident conditions. The changes do not introduce a new mode of
plant operation. While these changes may afford North Atlantic
operational flexibility, the changes are an enhancement and do not
affect plant safety. The plant's design and design basis are not
revised and the current safety analyses remains in effect.
Thus, these proposed revisions to TS Index, TS 3/4.9.3, TS 3/
4.9.4, and TS 3/4.9.9 do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The proposed changes to TS Index, TS 3/4.9.3, TS 3/4.9.4, and TS
3/4.9.9 do not adversely affect the safety margins established
through Limiting Conditions for Operation, Limiting Safety System
Settings and Safety Limits as specified in the Technical
Specifications nor is the plant design revised by the proposed
changes. The current licensing basis allows penetration isolation by
manual or automatic means.
Though a portion of the proposed change to TS 3/4.9.4 appears to
be a relaxation to the current licensing basis, North Atlantic has
incorporated administrative conservatism into TS 3/4.9.4 to ensure
the proposed changes, in conjunction with other TS required
surveillance testing, offset any potential minimal reduction in the
margin of safety. North Atlantic believes that the proposed change
to TS 3/4.9.4 is more conservative than that currently allowed in
the improved STS, NUREG-1431, Revision 2.
Thus, it is concluded that these proposed revisions to TS Index,
TS 3/4.9.3, TS 3/4.9.4, and TS 3/4.9.9 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Section Chief: James W. Clifford.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: August 15, 2001.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to (1) reflect the
replacement of Monticello's licensed operator initial and
requalification training programs with an accredited systems approach
to training program and (2) relocate the existing TS requirements for
procedures, records, and reviews to the operational quality assurance
plan.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes are administrative in nature and compliance
with applicable regulatory requirements will continue to be
maintained. The proposed changes do not involve any change to the
configuration or alter existing system relationships. In addition,
the proposed changes do not alter the conditions or assumptions in
any of the previous accident analyses thus, the radiological
consequences previously evaluated are not adversely affected by the
proposed changes.
Therefore, the probability or consequences of an accident
previously evaluated are not affected by the proposed amendment.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any previously analyzed.
The proposed changes are administrative in nature and compliance
with applicable regulatory requirements will continue to be
maintained. The proposed changes do not involve any change to the
configuration or method of operation of any plant equipment.
Accordingly, no new failure modes have been introduced for any plant
system or component important to safety nor has any new limiting
single failure been identified as a result of the proposed changes.
Also, there
[[Page 48291]]
will be no changes in types or increases in the amounts of any
effluents released offsite.
Therefore, the possibility of a new or different kind of
accident from any accident previously evaluated will not be created.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
The proposed changes are administrative in nature and do not
involve any change in the methodology or method of operation of any
plant equipment. The proposed changes do not involve any change to
the configuration or alter existing system relationships. The
appropriate controls to provide continued assurance of compliance to
applicable regulatory requirements has been maintained.
Therefore, the proposed amendment will not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Section Chief: Claudia M. Craig.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: August 31, 2001.
Description of amendment request: The proposed amendment would
amend the licenses to change the required implementation date for
previously issued Amendment No. 184 to Facility Operating License NPF-
14 and Amendment No. 158 to Facility Operating License NPF-22. The
proposed amendment would not alter any of the requirements of the
Susquehanna Steam Electric Station (SSES) Unit 1 and 2 Technical
Specifications (TSs). The previously issued amendments incorporate
long-term power stability solution instrumentation into the SSES Unit 1
and 2 TSs. When implemented, these amendments will incorporate into the
TSs the licensee's final response to GL 94-02, ``Long Term Solutions
and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic
Instabilities in Boiling Water Reactors.'' Specifically, these
amendments will, in part, add TS requirements related to the operating
power range monitoring (OPRM) system. The licensee stated that recently
identified deficiencies in the OPRM trip setpoint methodology, as
documented in a General Electric 10 CFR part 21 report issued on June
29, 2001, have adversely affected its ability to implement the subject
amendments. Therefore, the licensee requested that the required
implementation date for Amendment No. 184 to License No. NPF-14 and
Amendment No. 158 to License No. NPF-22 be revised to become effective
no later than November 1, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed amendment implementation date extension is
administrative in nature and does not require any physical plant
modifications, physically affect any plant systems or components, or
entail changes in plant operation. The resulting consequences of
transients and accidents will remain within the NRC approved
criteria. Therefore, the proposed action does not involve an
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 amendment implementation date extension is
administrative in nature and does not require any physical plant
modifications, physically affect any plant systems or components, or
entail changes in plant operation. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed amendment implementation date extension is
administrative in nature and does not require any physical plant
modifications, physically affect any plant systems or components,
nor entail changes in plant operation. Since the proposed changes do
not affect the physical plant or have any impact on plant operation,
the proposed changes will not jeopardize or degrade the function or
operation of any plant system or component. Therefore, the proposed
change does not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: Peter Tam, Acting.
Tennessee Valley Authority (TVA), Docket Nos. 50-260 and 50-296, Browns
Ferry Nuclear Plant (BFN), Units 2 and 3, Limestone County, Alabama
Date of amendment request: August 17, 2001.
Description of amendment request: The proposed amendments would
revise the reactor vessel pressure-temperature (P-T) limits depicted in
Technical Specification Figure 3.4.9-1 for each unit. In addition,
pursuant to 10 CFR 50.12, TVA is requesting an exemption from the
requirements of 10 CFR part 50, Appendix G, to allow the use of
American Society of Mechanical Engineers (ASME) Code Case N-640 as a
basis for these revised curves. Code Case N-640, ``Alternative
Requirement Fracture Toughness for Development of P-T Limit Curves for
ASME Boiler and Pressure Vessel Code Section XI, Division 1,'' permits
the use of the plane 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.
The exemption request is being reviewed separately.
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:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed Units 2 and 3 change deals exclusively with the
reactor vessel pressure-temperature (P-T) curves which define the
permissible regions for operation and testing. Failure of the
reactor vessel is not considered as a design basis accident. Through
the design conservatisms used to calculate the P-T curves, reactor
vessel failure has a low probability of occurrence and is not
considered in the safety analyses. The proposed changes adjust the
reference temperature for the limiting material to account for
irradiation effects and provide the same level of protection as
previously evaluated and approved. The adjusted reference
temperature calculations were performed using the guidance contained
in Regulatory Guide 1.99, Revision 2, and ASME Section XI Code Case
N-640 to reflect use of the operating limits to 19.5 Effective Full
Power Years (EFPY). These changes do not alter or prevent the
operation of equipment required to mitigate any accident analyzed in
the BFN Final Safety Analysis Report. Therefore, this change does
not increase the probability or consequences of any previously
evaluated accident.
[[Page 48292]]
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change to the Units 2 and 3 reactor vessel P-T
curves does not involve a modification to plant equipment. No new
failure modes are introduced. There is no effect on the function of
any plant system, and no new system interactions are introduced by
this change. Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed curves conform to the guidance contained in
Regulatory Guide 1.99, Revision 2, and maintain the safety margins
specified in 10 CFR 50, Appendix G. Therefore, the proposed
amendment does not involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1 (WBN), Rhea County, Tennessee
Date of amendment request: August 7, 2001 (TS-01-04).
Description of amendment request: The proposed amendment would add
a new condition and associated actions to the Technical Specification
Limiting Condition for Operation (LCO) 3.8.1, ``AC Sources Operating,''
to allow one Diesel Generator (DG) be out of service for 14 days.
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:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The emergency DGs are designed as backup AC power sources in the
event of loss of offsite power. The proposed AOT [allowed outage
time] does not change the conditions, operating configurations, or
minimum amount of operating equipment assumed in the safety analysis
for accident mitigation. No changes are proposed in the manner in
which the DGs provide plant protection or which create new modes of
plant operation. In addition, a Probabilistic Safety Analysis (PSA)
evaluation concluded that the risk contribution of the AOT extension
is non-risk significant. Therefore, the proposed amendment does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not introduce any new modes of plant
operation or make physical changes to plant systems. Therefore,
extension of the allowable AOT for DGs does not create the
possibility of a new or different accident.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The DGs are designed as backup AC power sources in the event of
loss of offsite power. The proposed AOT does not change the
conditions, operating configurations, or minimum amount of operating
equipment assumed in the safety analysis for accident mitigation. No
changes are proposed in the manner in which the DGs provide plant
protection or which create new modes of plant operation. In
addition, a PSA evaluation concluded that the risk contribution of
the AOT extension is non-risk significant. Therefore, the proposed
amendment does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: August 20, 2001.
Description of amendment request: The proposed change to the
Technical Specifications (TSs) would revise certain requirements
associated with demonstrating the operability of alternate trains when
redundant equipment is made or found to be inoperable. The TSs revised
include: 4.4.B, 4.5.A.2, 4.5.A.3, 4.5.A.4, 4.5.B.2, 4.5.C.2, 4.5.C.3,
4.5.D.2, 4.5.D.3, 4.5.E.2, 4.5.F.2, 4.5.H.1, 4.7.B.3.c, 4.10.B.1, and
4.10.B.3.b.2. Some format and typographical errors are also being
corrected.
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. Will the proposed changes involve a significant increase in
the probability or consequences of an accident previously evaluated?
Because changing surveillance test requirements does not change
the probability of accident precursors, this proposed change does
not affect the probability of an accident previously evaluated.
Since other periodic and post-maintenance surveillance requirements
ensure that the operability of systems and components is maintained,
there is no significant increase in the consequences of accidents
previously evaluated.
Furthermore, the removal of the additional surveillance testing
from the Technical Specifications would result in a decrease in the
probability of equipment failure because the excessive testing
causes unnecessary wear on the safety-related equipment and
unnecessary challenges to safety systems. Reduced testing may also
eliminate the potential for human error associated with system
alignments and misdirection of attention from monitoring and
directing plant operations.
Administrative changes to the Technical Specifications do not
alter any technical requirements, and as such, do not increase the
probability or consequences of accidents.
Therefore, the proposed change will not increase the probability
or consequences of any accident previously evaluated.
2. Will the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Reduced surveillance testing does not create new or different
kinds of accidents since modes of operation are unchanged and
additional accident precursors are not introduced. System
operability requirements and design bases remain the same, and
reactor operations are unchanged. Since system and component testing
only involves the assurance of operability, reduced testing does not
introduce mechanisms that may contribute to the possibility of new
or different kinds of accidents.
Administrative changes to the Technical Specifications do not
alter any technical requirements, and as such, do not create the
possibility of new or different kinds of accidents.
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. Will the proposed changes involve a significant reduction in
a margin of safety?
The proposed change will not decrease operability requirements,
nor reduce the equipment required during various plant conditions.
An acceptable level of testing exists in other Technical
Specification requirements to demonstrate system and component
operability. There are no changes to system or component operability
requirements; therefore, systems and
[[Page 48293]]
components will be available to provide existing margins of safety.
The same systems and components with the same performance levels
assumed in safety analyses will still be available to mitigate
consequences of postulated accidents.
Administrative changes to the Technical Specifications do not
alter any technical requirements, and as such, have no effect on
margins of safety.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
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, 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/NRC/ADAMS/index.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 email to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of application for amendment: March 29, 2001, as supplemented
by letters dated June 27, 2001, and July 24, 2001.
Brief description of amendment: The amendment revised the reactor
coolant system heatup, cooldown, and inservice leak hydrostatic test
limitations for the reactor coolant system to a maximum of 29 effective
full power years in accordance with Title 10 of the Code of Federal
Regulations, Part 50, Appendix G. These pressure-temperature (P-T)
limits are contained in TMI Unit 1 Technical Specification (TS) 3.1.2.
In addition, the amendment revised the low-temperature overpressure
protection (LTOP) requirements in TSs 3.1.12 and 4.5.2 to reflect the
revised P-T limits. These changes will allow operation of two reactor
coolant pumps in a single loop during LTOP conditions.
Date of issuance: September 6, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 234.
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 25, 2001 (66 FR
38758).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 6, 2001.
No significant hazards consideration comments received: No.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Docket No.
72-8, Calvert Cliffs Independent Spent Fuel Storage Installation,
Calvert County, Maryland
Date of application for amendments: November 22, 1999, as
supplemented by letters dated October 4 and November 10, 2000, and May
18, 2001.
Brief description of amendments: The amendments authorize revisions
to the Calvert Cliffs Nuclear Power Plant Updated Final Safety Analysis
Report and Independent Spent Fuel Storage Installation Updated Safety
Analysis Report to incorporate changes associated with the aircraft
hazards analysis due to increased ``random'' military flights in the
vicinity of these facilities. These changes constitute an unreviewed
safety question as defined in 10 CFR 50.59 and 10 CFR 72.48.
Date of issuance: August 29, 2001.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 246 and 221.
Renewed Facility Operating License Nos. DPR-53 and DPR-69 and
Materials License No. SNM-2502: Amendments revised licenses.
Date of initial notice in Federal Register: December 29, 1999 (64
FR 73085).
The supplemental letters dated October 4 and November 10, 2000, and
May 18, 2001, provided clarifying information that did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of these amendments is contained in
a Safety Evaluation dated August 29, 2001.
No significant hazards consideration comments received: No.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: December 11, 2000.
Brief description of amendment: The amendment revises the Technical
Specifications (TSs) to incorporate editorial revisions,
clarifications, and corrections. Specifically, the amendment: (1)
Provides updated information and corrections to the TS cover page,
table of contents, and list of figures, (2) revises TS 4.5.E, ``Control
Room Air Filtration System,'' to remove an incorrect system test
description and provide consistent test values for system flow rate and
filter efficiency, (3) revises TS 6.2.1.a, ``Facility Management and
[[Page 48294]]
Technical Support,'' to reference the Quality Assurance Program
Description as the location of the documentation rather than the
Updated Final Safety Analysis Report, (4) revises TS 6.9.1.7, ``Monthly
Operating Report,'' to change the recipient of the Monthly Operating
Report, and (5) corrects the periodicity of the Radioactive Effluent
Release Report from semi-annual to annual in TS 6.15, ``Offsite Dose
Calculation Manual'' and TS 6.16, ``Major Changes to Radioactive
Liquid, Gaseous and Solid Waste Systems.'' In addition, the amendment
revises TS Figure 5.1-1B concerning the indicated vent location
associated with Indian Point Unit 3 (IP3). The labels for the IP3 plant
vent and the machine shop were reversed.
Date of issuance: August 29, 2001.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 219.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 21, 2001 (66
FR 11057).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 29, 2001.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: April 23, 2001, as supplemented
June 25, June 29, and July 19, 2001.
Brief description of amendment: The amendment revises pressure-
temperature limit curves and cold overpressure protection limits.
Date of issuance: August 27, 2001.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance. August 27, 2001.
Amendment No.: 197.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 11, 2001 (66 FR
36340).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 27, 2001.
No significant hazards consideration comments received: No.
Exelon Generation Company, PSEG Nuclear LLC, and Atlantic City Electric
Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power
Station (PBAPS), Units 2 and 3, York County, Pennsylvania
Date of application for amendments: April 3, 2001.
Brief description of amendments: The amendments revised the PBAPS
Units 2 and 3 Technical Specifications (TSs) to incorporate Technical
Specification Task Force (TSTF) Item 258, Revision 4. TSTFs are changes
to the improved standard TS that were initiated by the nuclear power
industry and submitted to the NRC staff. TSTF-258, Revision 4, revises
TS Section 5.0, Administrative Controls, to delete specific TS staffing
requirements for licensed Reactor Operators (ROs) and Senior Reactor
Operators (SROs), relocate the working hour limits to a plant
procedure, clarify requirements for the Shift Technical Advisor
position, add regulatory definitions for ROs and SROs, revise the
Radioactive Effluent Controls Program to be consistent with the intent
of 10 CFR Part 20, and revises radiological area control requirements
for high radiation areas to be consistent with 10 CFR 20.1601(c).
Date of issuance: August 30, 2001.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendments Nos.: 240 and 243.
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 12, 2001 (66 FR
31708).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 30, 2001.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: March 7, 2001, as supplemented
April 25, June 20, and July 16, 2001.
Brief description of amendment: The amendment revised the Improved
Technical Specifications (ITS) 5.6.2.20, ``Containment Leakage Rate
Testing Program'' to allow a one-time interval increase for the Type A
Integrated Leakage Rate Test for no more than 5 years.
Date of issuance: August 30, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 197.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 2, 2001 (66 FR
17967). The supplemental letters provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 30, 2001.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit 2, Berrien County, Michigan
Date of application for amendments: September 1, 2000.
Brief description of amendments: The amendment approves changes to
the Updated Final Safety Analysis Report (UFSAR) regarding the modeling
of the pressurizer heater operation and spray effectiveness as they
relate to certain transients that are analyzed for pressurizer
overfill. Specifically, the amendment approves a change to the
moderator temperature coefficient currently in the UFSAR assumed as an
initial condition for the loss of all nonemergency alternating current
power and loss of normal feedwater transients.
Date of issuance: August 23, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 237.
Facility Operating License No. DPR-74: Amendment revised the
Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: September 20, 2000 (65
FR 56953).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 23, 2001.
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: January 18, 2001, as
supplemented April 20, 2001.
Brief description of amendment: The amendment revises the Kewaunee
Nuclear Power Plant (KNPP) Technical Specifications (TSs) 3.10.m to
increase the minimum reactor coolant flow from
[[Page 48295]]
85,500 gallons per minute (gpm) flow per loop to 93,000 gpm flow per
loop.
Date of issuance: September 5, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 157.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 21, 2001 (66
FR 11062).
The April 20, 2001, supplemental information 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 September 5, 2001.
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: June 18, 2001.
Brief description of amendment: The amendment deleted items 3 and 4
from Section 5.15, ``Post-Accident Radiological Sampling and
Monitoring,'' of the Fort Calhoun Station, Unit No. 1 Technical
Specifications, and thereby eliminates the requirements to have and
maintain the post-accident sampling system (PASS).
Date of issuance: August 29, 2001.
Effective date: August 29, 2001, and shall be implemented within
120 days from the date of issuance.
Amendment No.: 200.
Facility Operating License No. DPR-40. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 25, 2001 (66 FR
38765).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 29, 2001.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: April 11, 2001, as supplemented
June 13, 2001.
Brief description of amendment: The amendment revises the Hope
Creek Technical Specifications (TSs) to relax the frequency for testing
of excess flow check valves (EFCVs). Specifically, TS surveillance
requirement 4.6.3.4 has been changed to revise required testing of
EFCVs from once per 18 months for all valves to a test of a
representative sample each 18 months such that all valves are tested
once in 10 years.
Date of issuance: August 28, 2001.
Effective date: As of the date of issuance, and shall be
implemented during Refueling Outage 10, currently scheduled to commence
in October 2001.
Amendment No.: 132.
Facility Operating License No. NPF-57: This amendment revised the
TSs.
Date of initial notice in Federal Register: May 30, 2001 (66 FR
29361).
The June 13, 2001, letter 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 August 28, 2001.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: September 22, 2000.
Brief Description of amendments: These amendments revise the
Facility Operating Licenses ( FOLs) and the Technical Specifications
(TS) to remove obsolete license conditions, make editorial changes in
the FOLs, and implement associated changes to the TS and Bases.
Date of issuance: August 30, 2001.
Effective date: August 30, 2001.
Amendment Nos.: 227 and 227.
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
change the License and Technical Specifications.
Date of initial notice in Federal Register: November 1, 2000 (65
FR 65351).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 30, 2001.
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: March 22, 2001.
Brief description of amendment: The amendment (1) decreases the
allowable values for Function 8, pressurizer pressure-low and
pressurizer pressure-high, in Table 3.3.1-1, ``Reactor Trip System
Instrumentation,'' and (2) increases the allowable value for Function
1.d, pressurizer pressure-low for safety injection, in Table 3.3.2-1,
``Engineered Safety Feature Actuation System Instrumentation.''
Date of issuance: August 30, 2001.
Effective date: August 30, 2001, and shall be implemented prior to
entry into Mode 3 in the restart from refueling outage 12 scheduled for
the Spring 2002.
Amendment No.: 140.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 2, 2001 (66 FR
22035).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 30, 2001.
No significant hazards consideration comments received: No.
Note: The publication date for this notice will change from
every other Wednesday to every other Tuesday, effective January 8,
2002. The notice will contain the same information and will continue
to be published biweekly.
Dated at Rockville, Maryland, this 10th day of September, 2001.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 01-23209 Filed 9-18-01; 8:45 am]
BILLING CODE 7590-01-P