[Federal Register Volume 66, Number 247 (Wednesday, December 26, 2001)]
[Notices]
[Pages 66461-66477]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-31473]
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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 December 3, 2001 through December 14, 2001.
The last biweekly notice was published on December 12, 2001 (66 FR
64284).
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
[[Page 66462]]
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission 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. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By January 25, 2002, 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 NRC's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852.
Publicly available records will be accessible electronically 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 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,
[[Page 66463]]
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 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].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: April 17, 2001.
Description of amendment request: The proposed amendment would make
editorial and administrative corrections to Technical Specifications
(TS) Section 3.3, ``Instrumentation'', and eliminate minor
discrepancies between TS Section 3.3 and other plant licensing basis
documents.
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:
Does the Change Involve a Significant Increase in the Probability or
Consequences of an Accident Previously Evaluated?
The proposed changes involve correction of editorial or
administrative errors made during the conversion of the Clinton
Power Station (CPS) Technical Specifications (TS) to the improved TS
(ITS). These proposed changes are based upon current design and
licensing basis requirements. The proposed changes involve
correction or reformatting of the TS and do not involve any physical
changes to plant systems, including those that mitigate the
consequences of accidents or the manner in which these plant systems
are operated. As such, these changes do not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
Does the Change Create the Possibility of a New or Different Kind of
Accident From Any Accident Previously Evaluated?
The proposed changes involve correcting errors or reformatting
existing TS requirements that do not involve a physical alteration
of the plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. These changes are consistent with the assumptions in the
safety analyses and licensing basis. Thus, these changes do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
Does the Change Involve a Significant Reduction in a Margin of Safety?
The proposed changes involve correcting editorial or
administrative errors introduced during the conversion of the CPS TS
to the ITS. The change to the Allowable Value for the Control Room
Ventilation System air intake radiation monitors setpoint in TS
Table 3.3.7.1-1 is consistent with the supporting analyses for the
trip setpoint value that was previously contained in the TS. The
changes involve reformatting or correction of errors, and therefore
will not reduce any margin of safety because there is no effect on
any safety analysis assumptions. These proposed changes maintain
requirements within the safety analyses and licensing basis.
Therefore, these changes do not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Robert Helfrich, Mid-West Regional Operating
Group, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville,
IL 60555.
NRC Section Chief: Anthony J. Mendiola.
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: May 21, 2001.
Description of amendment request: The proposed amendment would
revise the actions required if the refueling equipment interlocks
become inoperable. The proposed changes are consistent with the changes
submitted to the Nuclear Regulatory Commission by the Technical
Specifications Task Force, Issue number 225, Revision 1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
Does the Change Involve a Significant Increase in the Probability or
Consequences of an Accident Previously Evaluated?
The proposed addition of alternate actions in the event that the
refueling equipment interlocks are determined to be in operable
ensures that the safety function provided by the interlocks are
enforced. This is accomplished through manually inserting a rod
block to prevent the inadvertent withdrawal of a control rod when
fuel is being moved over the core region.
The refueling equipment interlocks are credited in the Control
Rod Removal Error During Refueling--Fuel Insertion with Control Rod
Withdrawn as described in Updated Safety Analysis (USAR Section
15.4.1.1.2.2). The manual insertion of a control rod withdrawal
block provides equivalent protection for the conditional rod block
provided by the refueling equipment interlocks.
The proposed change to the surveillance frequency does not
change the means in which the refueling equipment operates. A review
of surveillance history was performed for the past two refueling
outages. In the last seven performances of the refueling equipment
interlocks operability test, the interlocks have operated
successfully with no corrective maintenance or corrective action
necessary. Therefore, since the proposed changes do not result in
any physical changes to the facility, or involve any modifications
to plant systems or design parameters or conditions that contribute
to the initiation of any accidents previously evaluated, the
proposed changes do not increase the probability of any accident
previously evaluated.
Since the proposed changes maintain the same level of protection
provided by the refueling equipment interlocks, the conclusion of
the accident scenario remain valid. The probability of a criticality
event during refueling remains such that no radioactive material
would be released. Therefore, the proposed changes do not increase
the consequences of an accident previously evaluated.
In summary, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Does the Change Create the Possibility of a New or Different Kind of
Accident From Any Accident Previously Evaluated?
The proposed changes do not involve a change to the plant design
or operation. Inserting a manual rod block is not considered an
abnormal operation. The change to the SR [surveillance requirement]
frequency does not increase the probability of a malfunction of the
refueling equipment interlocks, since the interlocks are considered
reliable and their function can be verified with each fuel move. As
a result, the proposed changes do not affect any of the parameters
or conditions that could contribute to the initiation of any
accidents. No new accident modes or equipment failure modes are
created by these changes.
[[Page 66464]]
Therefore, these proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Does the Change Involve a Significant Reduction in a Margin of Safety?
The major challenge to the margin of safety would be a
criticality event that would cause a potential failure of the fuel
cladding. The proposed addition of alternative actions in the event
that the refueling equipment interlocks are determined to be
inoperable ensure that equivalent protection is in place during fuel
loading movements. Given this equivalent protection, a criticality
event is not credible. In addition, the increase in the SR frequency
for performing the channel functional test of the refueling
equipment interlocks does not impact the ability of the interlocks
to perform their function, thereby maintaining the refueling
interlocks function.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Robert Helfrich, Mid-West Regional Operating
Group, Exelon Generation Company, LLC, 4300 Windfield Road,
Warrenville, IL 60555.
NRC Section Chief: Anthony J. Mendiola.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: December 7, 2001.
Description of amendment request: The proposed amendments would
allow implementation of 10 CFR Part 50, Appendix J, Option B, which
governs performance-based containment leakage testing requirements, for
Type B and C testing. In addition, the licensee also proposes to (a)
modify Technical Specification (TS) 3.6.3 to delete the requirement for
conducting soap bubble tests of welded penetrations during Type A tests
which are not individually Type B or Type C testable, and (b) to modify
TS 3.6.3 to delete a separate requirement for leak testing containment
purge lower and upper compartment and instrument room valves with
resilient seals. These valves will be covered by the overall
Containment Leakage Rate Testing Program. Associated changes to the
Bases are also proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The following discussion is a summary of the evaluation of the
changes contained in this proposed amendment against the 10 CFR
50.92(c) requirements to demonstrate that all three standards are
satisfied. A no significant hazards consideration is indicated if
operation of the facility in accordance with the proposed amendment
would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated, or
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated, or
3. Involve a significant reduction in a margin of safety.
First Standard
The proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Implementation of these changes will provide continued
assurance that specified parameters associated with containment
integrity will remain within acceptance limits as delineated in 10
CFR 50, Appendix J, Option B. The changes are consistent with
current safety analyses. Although some of the proposed changes
represent minor relaxation to existing TS requirements, they are
consistent with the requirements specified by Option B of 10 CFR 50,
Appendix J. The systems affecting containment integrity related to
this proposed amendment request are not assumed in any safety
analyses to initiate any accident sequence. Therefore, the
probability of any accident previously evaluated is not increased by
this proposed amendment. The proposed changes maintain an equivalent
level of reliability and availability for all affected systems. In
addition, maintaining leakage within analyzed limits assumed in
accident analyses does not adversely affect either onsite or offsite
dose consequences. Therefore, the proposed amendment does not
increase the consequences of any accident previously evaluated.
Second Standard
The proposed amendment will not create the possibility of a new
or different kind of accident from any accident previously
evaluated. No changes are being proposed which will introduce any
physical changes to the existing plant design. The proposed changes
are consistent with the current safety analyses. Some of the changes
may involve revision in the testing of components; however, these
are in accordance with the McGuire's current safety analyses and
provide for appropriate testing or surveillance that is consistent
with 10 CFR 50, Appendix J, Option B. The proposed changes will not
introduce new failure mechanisms beyond those already considered in
the current safety analyses. No new modes of operation are
introduced by the proposed changes. The proposed changes maintain,
at minimum, the present level of operability of any system that
affects containment integrity.
Third Standard
The proposed amendment will not involve a significant reduction
in a margin of safety. The provisions specified in Option B of 10
CFR 50, Appendix J allow changes to Type B and Type C test intervals
based upon the performance of past leak rate tests. 10 CFR 50,
Appendix J, Option B allows longer intervals between leakage tests
based on performance trends, but does not relax the leakage
acceptance criteria. Changing test intervals from those currently
provided in the TS to those provided in 10 CFR 50, Appendix J,
Option B does not increase any risks above and beyond those that the
NRC has deemed acceptable for the performance based option. In
addition, there are risk reduction benefits associated with
reduction in component cycling, stress, and wear associated with
increased test intervals. The proposed changes provide continued
assurance of leakage integrity of containment without adversely
affecting the public health and safety and will not significantly
reduce existing safety margins. Similar proposed changes have been
previously reviewed and approved by the NRC, and they are applicable
to McGuire.
Based upon the preceding discussion, Duke Energy has concluded
that the proposed amendment does not involve a significant hazards
consideration.
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. Albert Carr, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina.
NRC Section Chief: Richard J. Laufer, Acting.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: November 15, 2001.
Description of amendment request: Entergy Operations, Inc. is
proposing that the Grand Gulf Nuclear Station (GGNS) Operating License
be amended to revise the GGNS Technical Specification Surveillance
Requirements (SRs) pertaining to testing of the standby emergency
diesel generators (DGs) to allow DG testing during reactor operation.
The proposed change would remove the restriction associated with these
SRs that prohibits conducting the required testing of the DGs during
reactor operating Modes 1, 2, or 3.
Basis for proposed no significant hazards consideration
determination:
[[Page 66465]]
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 any accident previously
evaluated.
The DGs and their associated emergency loads are accident
mitigating features, not accident initiating equipment. Therefore,
there will be no impact on any accident probabilities by the
approval of the requested amendment.
The design of plant equipment is not being modified by these
proposed changes. As such, the ability of the DGs to respond to a
design basis accident will not be adversely impacted by these
proposed changes. The capability of the DGs to supply power in a
timely manner will not be compromised by permitting performance of
DG testing during periods of power operation. Additionally, limiting
testing to only one DG at a time ensures that design basis
requirements for backup power is met, should a fault occur on the
tested DG. Therefore, there would be no significant impact on any
accident consequences.
Based on the above, the proposed change to permit certain DG
surveillance tests to be performed during plant operation will have
no effect on accident probabilities or consequences.
(2) The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
No new accident causal mechanisms would be created as a result
of NRC [Nuclear Regulatory Commission] approval of this amendment
request since no changes are being made to the plant that would
introduce any new accident causal mechanisms. Equipment will be
operated in the same configuration with the exception of the plant
mode in which the testing is conducted. This amendment request does
not impact any plant systems that are accident initiators; neither
does it adversely impact any accident mitigating systems.
Based on the above, implementation of the proposed changes would
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
(3) The proposed change does not involve a significant reduction
in the margin of safety.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their design functions
during and following an accident situation. These barriers include
the fuel cladding, the reactor coolant system, and the containment
system. The proposed changes to the testing requirements for the
plant DGs do not affect the operability requirements for the DGs, as
verification of such operability will continue to be performed as
required (except during different allowed Modes).
Continued verification of operability supports the capability of
the DGs to perform their required function of providing emergency
power to plant equipment that supports or constitutes the fission
product barriers. Consequently, the performance of these fission
product barriers will not be impacted by implementation of this
proposed amendment.
In addition, the proposed changes involve no changes to
setpoints or limits established or assumed by the accident analysis.
On this and the above basis, no safety margins will be impacted.
Therefore, implementation of the proposed changes would not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., 12th Floor, 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: October 29, 2001.
Description of amendment request: The proposed amendments would
revise technical specification (TS) 3.9.3, ``Refueling Operations--
Decay Time,'' by reducing the amount of time that the reactor must be
subcritical before the licensee is allowed to move irradiated fuel
assemblies in the reactor pressure vessel from 150 hours to 100 hours.
The amendment also makes various editorial, format and administrative
changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed change does not alter the manner in which fuel
assemblies are handled or core alterations are performed. The
proposed change does not alter the manner in which heavy loads are
controlled at BVPS. The proposed change does not result in changes
being made to structures, systems, or components (SSCs), or to event
initiators or precursors. Also, the proposed change does not impact
the design of plant systems such that previously analyzed SSCs would
now be more likely to fail. The initiating conditions and
assumptions for accidents described in the Updated Final Safety
Analysis Report (UFSAR) remain as previously analyzed. Thus, the
proposed change does not involve a significant increase in the
probability of an accident previously evaluated.
The proposed revision of the decay time from 150 hours to 100
hours is consistent with the assumptions used in the NRC approved
fuel handling accident (FHA) analyses for Beaver Valley Power
Station (BVPS) Unit Nos. 1 and 2. The BVPS radiological analyses
demonstrates that should a FHA occur within the containment or the
fuel building that involves irradiated fuel with at least 100 hours
of decay, the projected offsite doses for this event will be well
within the applicable regulatory limits.
Limiting Condition for Operation (LCO) 3.9.3, ``Refueling
Operations--Decay Time,'' will continue to ensure that irradiated
fuel is not moved in the reactor pressure vessel until at least 100
hours after shutdown which is consistent with the FHA radiological
analysis. This LCO will continue to ensure that key assumptions used
in the radiological safety analysis are met. The previously analyzed
SSCs are unaffected by the proposed change and continue to provide
assurance that they are capable of performing their intended design
function in mitigating the effects of design basis accidents (DBAs).
As such, the consequences of accidents previously evaluated in the
UFSAR will not be increased and no additional radiological source
terms are generated. Therefore, there will be no reduction in the
capability of those SSCs in limiting the radiological consequences
of previously evaluated accidents and reasonable assurance that
there is no undue risk to the health and safety of the public will
continue to be provided. Thus, the proposed change does not involve
a significant increase in the consequences of an accident previously
evaluated.
The proposed administrative, editorial, and format changes do
not affect the probability or consequences of any accident.
Therefore, the proposed amendment does not significantly
increase the probability or consequences of any accident previously
evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. The proposed amendment does not affect a previously
evaluated accident; e.g., FHA. The proposed amendment takes credit
for the normal decay of irradiated fuel and the existing
radiological analyses for FHAs.
The proposed change does not involve physical changes to
analyzed SSCs or changes to the modes of plant operation defined in
the technical specification. The proposed change does not involve
the addition or modification of plant equipment (no new or different
type of equipment will be installed) nor does it alter the design or
operation of any plant systems. No new accident scenarios, accident
or transient initiators or precursors, failure mechanisms, or
limiting single failures are introduced as a result of the proposed
change.
The proposed change does not cause the malfunction of safety-
related equipment assumed to be operable in accident analyses. No
new or different mode of failure has been created and no new or
different equipment performance requirements are imposed for
[[Page 66466]]
accident mitigation. As such, the proposed change has no effect on
previously evaluated accidents.
The proposed administrative, editorial, and format changes do
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
No. The proposed revision of the decay time from 150 hours to
100 hours is consistent with the assumptions used in the NRC
approved FHA accident analyses for BVPS Unit Nos. 1 and 2 and thus
does not involve a significant reduction in a margin of safety.
The proposed amendment does not alter the manner in which fuel
assemblies are handled or core alterations are performed. The
proposed amendment does not alter the manner in which heavy loads
are controlled at BVPS.
The proposed changes to the TS requirements will continue to
ensure that the necessary plant equipment is operable in the plant
conditions where these systems are required to operate to mitigate a
DBA. The proposed administrative, editorial, and format changes do
not affect plant safety.
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: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: L. Raghavan, Acting.
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: October 31, 2001.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TSs) by relocating the pressure
temperature Limit Curves and Low Temperature Overpressure Protection
(LTOP) and by creating a Pressure-Temperature Limits Report in
accordance with Generic Letter 96-03 (GL-96-03), ``Relocation of the
Pressure Temperature Limit Curves and Low Temperature Overpressure
Protection System Limits.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed changes are a relocation of the Reactor Coolant
System (RCS) pressure/temperature (P/T) limits, overpressure
protection system (OPPS) setpoint, and the enable temperature from
the Technical Specifications to the proposed Pressure and
Temperature Limits Report (PTLR). The PTLR is created in accordance
with the guidance provided by Generic Letter (GL) 96-03 and is
consistent with the content of NUREG-1431. The RCS P/T limits, OPPS
setpoint, and enable temperature will continue to meet the
requirements of 10 CFR 50, Appendix G, and will be generated in
accordance with the NRC approved methodology described in WCAP-
14040-NP-A, Rev. 2 with the exceptions noted in Technical
Specification Section 6.9.6.
Since the proposed changes are administrative in nature and do
not involve any change to any values being relocated, the proposed
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. As stated above, the proposed changes to relocate the RCS P/
T limits, OPPS setpoint, and the enable temperature from the
Technical Specifications to the PTLR are administrative changes. The
proposed changes do not result in a physical change to the plant or
add any new or different operating requirements on plant systems,
structures, or components.
Therefore, the proposed changes do not result in a significant
increase in the possibility of a new or different accident from any
previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
No. The margin of safety is not affected by the creation of the
proposed PTLR. Operation of the plant in accordance with the limits
specified in the PTLR will continue to meet the requirements of 10
CFR 50, Appendix G, with the identified exceptions, and will assure
that a margin of safety is not significantly decreased as the result
of the proposed changes.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Lakshminaras Raghavan (Acting).
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station (DBNPS), Unit 1, Ottawa County, Ohio
Date of amendment request: October 9, 2001.
Description of amendment request: The proposed amendment changes
affected Technical Specifications (TS) 3/4.3.2.2, ``Instrumentation--
Steam and Feedwater Rupture Control System Instrumentation,'' including
Table 3.3-11, ``Steam and Feedwater Rupture Control System
Instrumentation,'' Table 3.3-12, ``Steam and Feedwater Rupture Control
System Instrumentation Trip Setpoints,'' and Table 4.3-11 ``Steam and
Feedwater Rupture Control System Instrumentation Surveillance
Requirements.'' Related administrative changes are proposed to TS 3/
4.3.2.3, ``Instrumentation--Anticipatory Reactor Trip System
Instrumentation,'' Table 3.3-17, ``Anticipatory Reactor Trip System
Instrumentation,'' and TS 3/4.3.3.1, ``Instrumentation--Monitoring
Instrumentation--Radiation Monitoring Instrumentation,'' Table 3.3-6,
``Radiation Monitoring Instrumentation.'' Related changes to associated
TS Bases 3/4.3.1 and 3/4.3.2, ``Reactor Protection System and Safety
System Instrumentation,'' are also proposed.
The main purpose for this license amendment request is to decrease
the channel functional test frequency from monthly to quarterly for the
Steam and Feedwater Rupture Control System (SFRCS) Instrumentation
Channels.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below. These changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because the proposed changes do not
change any accident initiator, initiating condition, or assumption.
The proposed changes would revise Technical Specification (TS)
Table 3.3-11, ``Steam and Feedwater Rupture Control System
Instrumentation,'' and Table 4.3-11 ``Steam and Feedwater Rupture
Control System Instrumentation Surveillance Requirements,'' to
identify the Steam and Feedwater Rupture Control System (SFRCS)
output logic as a separate Functional Unit. In addition, the
proposed changes would revise TS Table 3.3-12, ``Steam and Feedwater
[[Page 66467]]
Rupture Control System Instrumentation Trip Setpoints,'' to remove
the ``Trip Setpoint'' values and also modify the ``Allowable
Values'' entry for Functional Unit 3, ``Steam Generator Feedwater
Differential Pressure--High,'' consistent with updated calculations
and current setpoint methodology, and revise the applicability of TS
Allowable Values for other SFRCS Functional Units in this table. The
proposed changes would also revise TS Table 4.3-11 to change the
Channel Functional Test surveillance requirements for the SFRCS
instrument channels from monthly to quarterly, consistent with
current methodology. The proposed changes would also make related
administrative changes to TS Limiting Condition for Operation (LCO)
3.3.2.2, TS Table 3.3-17, ``Anticipatory Reactor Trip System
Instrumentation,'' TS Table 3.3-6, ``Radiation Monitoring
Instrumentation,'' and the associated TS Bases.
These proposed changes do not involve a significant change to
plant design or operation.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
invalidate assumptions used in evaluating the radiological
consequences of an accident, do not alter the source term or
containment isolation, and do not provide a new radiation release
path or alter radiological consequences.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because the proposed
changes do not introduce a new or different accident initiator or
introduce a new or different equipment failure mode or mechanism.
3. Not involve a significant reduction in a margin of safety as
defined in the basis for any Technical Specification. The SFRCS
instrumentation setpoint analyses will continue to adequately
preserve the margin of safety. In addition, there are no new or
significant changes to the initial conditions contributing to
accident severity or consequences. Therefore, there are no
significant reductions in a margin of safety.
Conclusion:
On the basis of the above, the Davis-Besse Nuclear Power Station
has determined that the License Amendment Request does not involve a
significant hazards consideration. As this License Amendment Request
concerns a proposed change to the Technical Specifications that must
be reviewed by the Nuclear Regulatory Commission, this License
Amendment Request does not constitute an unreviewed safety question.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station (DBNPS), Unit 1, Ottawa County, Ohio
Date of amendment request: October 12, 2001.
Description of amendment request: The proposed amendment would
change the Operating License (OL) paragraph 2.C(1), Maximum Power
Level; OL paragraph 2.C(3)(d), Additional Conditions; Technical
Specification (TS) 1.3, Definitions--Rated Thermal Power; TS 2.1.1,
Safety Limits--Reactor Core, and associated Bases; TS 2.2.1, Limiting
Safety System Settings--Reactor Protection System Setpoints, and
associated Bases; TS 3/4.1.1.3, Reactivity Control Systems--Moderator
Temperature Coefficient; TS 3/4.2.5, Power Distribution Limits--DNB
Parameters; TS 3/4.4.9.1, Reactor Coolant System--Pressure/Temperature
Limits, and associated Bases; and TS 6.9.1.7, Core Operating Limits
Report. The purpose of this license amendment application would make
the necessary revisions to the Davis-Besse Nuclear Power Station
(DBNPS) TS to reflect an increase in the authorized rated thermal power
from 2772 MWt to 2817 MWt (approximately 1.63 percent), based on the
use of Caldon Inc. Leading Edge Flow Meter (LEFM) CheckPlus\TM\ System
instrumentation to improve the accuracy of the feedwater mass flow
input to the plant power calorimetric measurement.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
1a. Not involve a significant increase in the probability of an
accident previously evaluated based on the comprehensive analytical
efforts that were performed to demonstrate the acceptability of the
proposed power uprate changes. The proposed changes include:
revision of the maximum power level limit stated in Operating
License (OL) paragraph 2.C(1) and Technical Specification (TS)
Section 1.3, increasing the allowable power level from 2772 MWt to
2817 MWt; revision of the reactor core safety limits specified in TS
Section 2.1.1; revision of the Reactor Protection System (RPS) high
flux and Reactor Coolant System (RCS) pressure-temperature setpoints
provided in TS Section 2.2.1; revision of the RCS pressure-
temperature limits in TS Section 3/4.4.9.1, and a related change to
OL paragraph 2.C(3)(d); and revision of administrative controls
associated with the Core Operating Limits Report, as described in TS
Section 6.9.1.7. In addition, related changes to the TS Bases
associated with these TS Sections are proposed. An evaluation has
been performed that identified the systems and components that could
be affected by these proposed changes. The evaluation determined
that these systems and components will function as designed and that
performance requirements remain acceptable.
The primary loop components (reactor vessel, reactor internals,
control rod drive mechanisms (CRDMs), loop piping and supports,
reactor coolant pumps, steam generators and pressurizer) will
continue to comply with their applicable structural limits and will
continue to perform their intended design functions. Thus, there is
no increase in the probability of a structural failure of these
components leading to an accident.
The Leak-Before-Break analysis conclusions remain valid and the
breaks previously exempted from structural consideration remain
unchanged.
All of the Nuclear Steam Supply System (NSSS) systems will
continue to perform their intended design functions during normal
and accident conditions. The pressurizer spray flow remains above
its design value. Thus, the control system design analyses, which
credit the flow, do not require any modification. The components
continue to comply with applicable structural limits and will
continue to perform their intended design functions. Thus, there is
no increase in the probability of a structural failure of these
components.
All of the NSSS/Balance of Plant (BOP) interface systems will
continue to perform their intended design functions. The main steam
safety valves will provide adequate relief capacity to maintain the
main steam system within design limits.
The current loss of coolant accident (LOCA) hydraulic forcing
functions remain bounding.
The reduction in power measurement uncertainty through the use
of the Caldon Leading Edge Flow Meter (LEFM) CheckPlusTM
system, allows for certain safety analyses to continue to be used,
without modification, at the 2827 MWt power level (102% of 2772
MWt). Other safety analyses performed at a nominal power level of
2772 MWt have been either re-performed or re-evaluated at the 2817
MWt power level, and continue to meet their applicable acceptance
criteria. Some existing safety analyses had been previously
performed at a power level greater than 2827 MWt, and thus continue
to bound the 2817 MWt power level.
The proposed changes to the RCS pressure-temperature limit
curves impose a conservative projection of the increase in neutron
fluence associated with the power uprate. This projection will
ensure that the requirements of 10 CFR 50 Appendix G, ``Fracture
Toughness Requirements,'' will continue to be met following the
proposed power uprate. The design basis events that were protected
against by these limits have not changed, therefore, the probability
of an accident previously evaluated is not increased.
In addition to the changes related to the proposed power uprate,
unrelated changes are proposed to revise the moderator temperature
coefficient requirements listed
[[Page 66468]]
in TS Section 3.1.1.3, and to revise requirements relating to the
Departure from Nucleate Boiling (DNB) parameters listed in TS
Section 3.2.5. These proposed changes are conservative changes and
clarifications that do not involve any physical change to systems or
components, nor do they alter the typical manner in which the
systems or components are operated. Therefore, these changes will
not result in a significant increase in the probability of an
accident.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed power uprate
changes do not alter any assumptions previously made in the
radiological consequence evaluations, nor affect mitigation of the
radiological consequences of an accident previously evaluated.
The accident radiation dose evaluation was performed at 2827 MWt
and is bounding when operating at the proposed 2817 MWt using the
LEFM CheckPlusTM flow instrumentation.
The proposed changes unrelated to the power uprate also do not
alter any assumption previously made in the radiological consequence
evaluations, nor do they affect mitigation of the radiological
consequences of an accident previously evaluated. Therefore, these
changes will not involve a significant increase in the consequences
of an accident previously evaluated.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because no new
accident scenarios, failure mechanisms or single failures are
introduced as a result of the proposed power uprate changes as well
as the proposed changes unrelated to the power uprate. All systems,
structures, and components previously required for the mitigation of
an event remain capable of fulfilling their intended design
function. The proposed changes have no adverse effects on any
safety-related system or component and do not challenge the
performance or integrity of any safety-related system.
3. Not involve a significant reduction in a margin of safety
because extensive analyses of the primary fission product barriers,
conducted in support of the proposed power uprate, have concluded
that all relevant design criteria remain satisfied, both from the
standpoint of the integrity of at the primary fission product
barrier and from the standpoint of compliance with the regulatory
acceptance criteria. As appropriate, all evaluations have been
performed using methods that have either been reviewed and approved
by the Nuclear Regulatory Commission (NRC) or that are in compliance
with applicable regulatory review guidance and standards. The
proposed changes unrelated to the power uprate do not involve a
significant reduction in a margin of safety because they do not
involve the potential for a significant increase in a failure rate
of any system or component, and existing system and component
redundancy is not affected. Also, these changes do not involve any
new or significant changes to the initial conditions contributing to
accident severity or consequences.
Conclusion:
On the basis of the above, the Davis-Besse Nuclear Power Station
has determined that the License Amendment Request does not involve a
significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: November 26, 2001.
Description of amendment request: The licensee proposed to amend
the Technical Specifications (TSs) to delete Section 3/4.2.6,
``Inservice Inspection and Testing,'' and its associated bases, revise
Section 4.2.7, ``Reactor Coolant System Isolation Valves,'' and its
associated bases, create a new Section 6.17, ``Inservice Testing
Program,'' and delete several reporting requirements in Section 6.9.3,
``Special Reports.'' These changes will improve the TSs, making it
consistent with current NRC guidance and the improved Standard
Technical Specifications for General Electric (GE) Boiling Water
Reactor (BWR)/4 and BWR/6 plants (NUREG-1433 and NUREG-1434,
respectively). Most of these changes would also render the TSs to be
similar to the Nine Mile Point Nuclear Station, Unit No. 2 TSs, which
is based on NUREG-1433 and NUREG-1434.
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 operation of Nine Mile Point Unit 1 in accordance with
the proposed amendment will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed amendment deletes duplicative and unnecessary
inservice inspection (ISI) and inservice testing (IST) requirements
from the Technical Specifications; clarifies remaining IST
requirements; revises a requirement to perform quarterly testing of
the reactor coolant isolation valves to conform to the periodic
testing requirements of the ASME [American Society of Mechanical
Engineers] Boiler and Pressure Vessel Code (ASME Code); and deletes
unnecessary reporting requirements relating to routine ISI, primary
containment leakage testing, and secondary containment leakage
testing. These changes do not reduce the plant's existing ISI/IST
commitments based on 10CFR50.55a, Section XI of the ASME Code, and
Generic Letter 88-01. These changes also do not involve hardware
changes, changes in plant setpoints, or changes in plant safety
parameters.
Based on the above, the operation of Nine Mile Point Unit 1
(NMP1) in accordance with the proposed amendment, will not involve a
significant increase in the probability or the consequences of an
accident previously evaluated.
2. The operation of Nine Mile Point Unit 1 in accordance with
the proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not involve any physical modifications
to the plant nor alter equipment configuration, setpoints, or safety
parameters. The ISI/IST related changes are consistent with current
NRC guidance and industry standards and will continue to ensure
acceptable equipment operability and availability.
Based on the above, the operation of NMP1 in accordance with the
proposed amendment cannot create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The operation of Nine Mile Point Unit 1 in accordance with
the proposed amendment will not involve a significant reduction in a
margin of safety.
The proposed changes do not affect any of the plant's fission
product barriers or safety/operational limits. The ISI/IST related
changes will continue to ensure acceptable equipment operability and
availability.
Based on the above, the operation of NMP1 in accordance with the
proposed amendment will not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendment involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: L. Raghavan, Acting.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station Unit No. 2, Oswego County, New York
Date of amendment request: November 20, 2001.
Description of amendment request: The licensee proposed to amend
the Technical Specifications (TSs) regarding the safety limit minimum
critical power
[[Page 66469]]
ratio (SLMCPR) to reflect the results of cycle-specific calculations
performed for the next fuel cycle (i.e., Cycle 9), using NRC-approved
methodology for determining SLMCPR values. The proposed amendment would
also editorially revise references to topical reports which document
the approved methodology.
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 operation of Nine Mile Point Unit 2, in accordance with
the proposed amendment, will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The derivation of the revised Safety Limit Minimum Critical
Power Ratio (SLMCPR) values for Nine Mile Point Unit 2 (NMP2) Cycle
9 for incorporation into the Technical Specifications (TS) and their
use to determine cycle-specific thermal limits has been performed
using the NRC-approved methods and procedures in [Topical Report]
NEDE-24011-P-A, ``General Electric Standard Application for Reactor
Fuel'' (GESTAR II). The analysis methodology incorporates cycle-
specific parameters and reduced power distribution uncertainties in
the determination of the SLMCPR values. These calculations do not
change the method of operating the plant and have no effect on the
probability of an accident initiating event or transient.
The basis of the Minimum Critical Power Ratio Safety Limit is to
ensure no mechanistic fuel damage is calculated to occur if the
limit is not violated. The new SLMCPR values preserve the existing
margin to transition boiling and the probability of fuel damage is
not increased. The deletion of listed documents that are already
incorporated by reference into GESTAR II is administrative only.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The operation of Nine Mile Point Unit 2, in accordance with
the proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The new SLMCPR values for the NMP2 Cycle 9 core reload have been
calculated in accordance with the methods and procedures described
in GESTAR II. These methods have been reviewed and approved by the
NRC. The deletion of listed documents that are already incorporated
by reference into GESTAR II is administrative only. The changes do
not involve any new method for operating the facility and do not
involve any facility modifications. No new initiating events or
transients result from these changes. Therefore, the proposed TS
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The operation of Nine Mile Point Unit 2, in accordance with
the proposed amendment, will not involve a significant reduction in
a margin of safety.
The margin of safety as defined in the TS bases will remain the
same. The new, cycle-specific SLMCPR values are calculated using
NRC-approved methods and procedures that are in accordance with the
current fuel design and licensing criteria. The SLMCPR values remain
high enough to ensure that greater than 99.9% of all fuel rods in
the core are expected to avoid transition boiling if the limits are
not violated, thereby preserving the fuel cladding integrity. The
deletion of listed documents that are already incorporated by
reference into GESTAR II is administrative only. Therefore, the
proposed TS changes do not involve a significant reduction in [a]
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendment involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: L. Raghavan, Acting.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: November 19, 2001.
Description of amendment request: A change is proposed to Technical
Specification (TS) 3.0.3 to allow a longer period of time to perform a
missed surveillance. The time is extended from the current limit of ``*
* * up to 24 hours or up to the limit of the specified Frequency,
whichever is less'' to ``* * * up to 24 hours or up to the limit of the
specified Frequency, whichever is greater.'' In addition, the following
requirement would be added to the specification: ``A risk evaluation
shall be performed for any Surveillance delayed greater than 24 hours
and the risk impact shall be managed.''
The Nuclear Regulatory Commission (NRC) staff issued a notice of
opportunity for comment in the Federal Register on June 14, 2001 (66 FR
32400), on possible amendments concerning missed surveillances,
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 September 28, 2001 (66 FR
49714). The licensee affirmed the applicability of the following NSHC
determination in its application dated November 7, 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 proposed change relaxes the time allowed to perform a missed
surveillance. The time between surveillances is not an initiator of
any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The
equipment being tested is still required to be operable and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected. Any reduction
in confidence that a standby system might fail to perform its safety
function due to a missed surveillance is small and would not, in the
absence of other unrelated failures, lead to an increase in
consequences beyond those estimated by existing analyses. The
addition of a requirement to assess and manage the risk introduced
by the missed surveillance will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
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. A
missed surveillance will not, in and of itself, introduce new
failure modes or effects and any increased chance that a standby
system might fail to perform its safety function due to a missed
surveillance would not, in the absence of other unrelated failures,
lead to an accident beyond those previously evaluated. The addition
of a requirement to assess and manage the risk introduced by the
missed surveillance will further minimize possible concerns. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The extended time allowed to perform a missed surveillance does
not result in a significant reduction in the margin of safety. As
supported by the historical data, the likely
[[Page 66470]]
outcome of any surveillance is verification that the LCO [Limiting
Condition for Operation] is met. Failure to perform a surveillance
within the prescribed frequency does not cause equipment to become
inoperable. The only effect of the additional time allowed to
perform a missed surveillance on the margin of safety is the
extension of the time until inoperable equipment is discovered to be
inoperable by the missed surveillance. However, given the rare
occurrence of inoperable equipment, and the rare occurrence of a
missed surveillance, a missed surveillance on inoperable equipment
would be very unlikely. This must be balanced against the real risk
of manipulating the plant equipment or condition to perform the
missed surveillance. In addition, parallel trains and alternate
equipment are typically available to perform the safety function of
the equipment not tested. Thus, there is confidence that the
equipment can perform its assumed safety function.
Therefore, this change does not involve a significant reduction
in a margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the 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: John H. O'Neill, Jr., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: William D. Reckley, Acting.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: February 2, 2001, supplemented August
31, 2001.
Description of amendment request: The proposed amendments would
revise the technical specifications (TSs) to clarify the plant
conditions under which various specifications are applicable. The
licensee stated in its amendment request that a literal reading of the
current technical specifications wording may result in situations where
a routine plant shutdown would seem to be prohibited by TSs and,
thereby, require entry into TS 3.0.C. This amendment request also makes
several administrative changes to the TSs, including revising
references to the Chief Nuclear Corporate Officer, capitalizing defined
terms, and updating references to previously relocated TS paragraphs
and correcting the List of Figures. The licensee's supplement to the
amendment request, dated August 31, 2001, proposed a correction of a
typographical error in TS Table 3.5-2B, Action 33.
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 operation of the facility with the proposed amendment
involve a significant increase in the probability or consequences of
an accident previously evaluated?
The proposed changes are administrative in nature and clarify
existing specifications without reducing or altering the
requirements imposed by existing specifications. The proposed
changes do not significantly affect any system that is a contributor
to initiating events for previously evaluated accidents. Neither do
the changes significantly affect any system that is used to mitigate
any previously evaluated accidents. Therefore, the proposed changes
do not involve any significant increase in the probability or
consequence of an accident previously evaluated.
2. Does operation of the facility with the proposed amendment
create the possibility of a new or different kind of accident from
any accident previously evaluated?
The proposed changes are administrative in nature and clarify
existing specifications without reducing or altering the
requirements imposed by existing specifications. The proposed
changes do not alter the design, function, or operation of any plant
component and do not install any new or different equipment,
therefore a possibility of a new or different kind of accident from
those previously analyzed has not be[en] created.
3. Does operation of the facility with the proposed amendment
involve a significant reduction in a margin of safety?
The proposed changes are administrative in nature and clarify
existing specifications without reducing or altering the requirements
imposed by existing specifications. Thus, the proposed change[s] do not
involve a significant reduction in the margin of safety associated with
the safety limits inherent in either the principle barriers to a
radiation release (fuel cladding, RCS [reactor coolant system]
boundary, and reactor containment), or the maintenance of critical
safety functions (subcriticality, core cooling, ultimate heat sink, RCS
inventory, RCS boundary integrity, and containment integrity).
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB,
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
NRC Acting Section Chief: William D. Reckley, Acting.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: November 21, 2001.
Description of amendment request: The proposed amendment will
revise Technical Specifications 2.15(5) and 2.15(6) to identify: (1)
all indication and control functions required for the alternate
(remote) shutdown panels, (2) panel locations of the functions, and (3)
the number of operable channels required.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to Technical Specifications Sections
2.15(5) and 2.15(6) identify functions, instruments, and controls
along with their location and the number of required channels. New
Technical Specifications Section 2.15(5) addresses the regulatory
requirements for equipment required for Alternative and Dedicated
Shutdown Capability per 10 CFR part 50, Appendix R. It will ensure
that proper Limiting Conditions for Operation are entered for
equipment or functional inoperability. There are no physical
alterations being made to the Alternate Shutdown Panels and the
Auxiliary Feedwater Panel or related systems. Therefore, the
proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes will not result in any physical alterations
to the Alternate Shutdown Panels or the Auxiliary Feedwater Panel,
or any plant configuration, systems, equipment, or operational
characteristics. There will be no changes in operating modes, or
safety limits, or instrument limits. With the proposed changes in
place, Technical Specifications retain requirements for the
Alternate Shutdown Panels and the Auxiliary Feedwater Panel.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes clarify the regulatory requirements for the
Alternative and Dedicated Shutdown Capability as defined by 10 CFR
Part 50, Appendix R. The proposed changes will not alter any
physical
[[Page 66471]]
or operational characteristics of the Alternate Shutdown Panels and
the Auxiliary Feedwater Panel and their associated systems and
equipment. Therefore, the proposed changes do not involve a
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: November 21, 2001.
Description of amendment request: The proposed amendment will add
three topical report references to Technical Specification 5.9.5,
``Core Operating Limit Reports.''
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 incorporates three additional Framatome
ANP topical reports for conducting core reload analyses. Since the
intent of the amendment request is to add references to NRC-approved
reload analysis methods, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
No new or different modes of operation are proposed as a result
of these changes. The proposed revision does not change any
equipment required to mitigate the consequences of an accident. The
proposed addition of NRC-approved topical reports to the Technical
Specification does not modify the manner in which the topical
reports may be implemented. The plant will continue to operate
within the limits specified by the Core Operating Limits Report and
corrective actions will be taken in accordance with the Technical
Specifications should these limits be exceeded. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
As required by Technical Specification 5.9.5, the analytical
methods used to determine the core operating limits shall be those
previously reviewed and approved by the NRC. The proposed change
incorporates methodologies applicable for use with fuel supplied by
Framatome ANP that have been approved by the NRC as documented by
Safety Evaluation Reports (References 10.1, 10.2, and 10.3 [of the
November 21, 2001, amendment request]). Because Technical
Specification 5.9.5 also requires that the core operating limits
shall be determined and requires that all applicable limits of the
safety analysis are met, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California
Date of amendment request: December 28, 2000, as supplemented by
letters dated March 29 and October 31, 2001.
Description of amendment request: The proposed amendment would
convert the Humboldt Bay Power Plant Unit 3 Current Technical
Specifications to a set of Permanently Defueled Technical
Specifications with a more standardized format and content based on a
revision to 10 CFR 50.36 (Technical Specifications) and technical
specifications approved for other permanently shutdown nuclear power
plants (Millstone Unit 1 and Trojan Nuclear Plant).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analyses of the issue of no significant hazards
consideration, which are presented below.
The conversion of the Humboldt Bay Power Plant (HBPP) Current
Technical Specifications (CTS) to Permanently Defueled Technical
Specifications (PDTS) involves the following four types of
dispositions:
A Administrative reformatting and rewording
D Item deleted from the Technical Specifications (TS)
LG Relocating items from CTS to the Defueled Safety Analysis Report
(DSAR), PDTS, or other Licensee-Controlled Document
N Addition of new requirements of new sections to the PDTS
Administrative Reformatting and Rewording
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change involves reformatting and editorially
rewording of the CTS. As such, this change is administrative in
nature and does not impact initiators of analyzed events or assumed
mitigation of accidents or transient events. Therefore, this change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change will not impose any different
operational requirements and any administrative additions are non-
operational in nature and have not been identified and justified.
Thus, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change will not reduce a margin of safety because
it has no impact on any safety analysis assumptions. This change is
administrative in nature. As such, no question of safety is
involved.
Items Deleted from the Technical Specifications that are Duplicative in
Nature to Other Regulatory Requirements
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change involves deleting information from the CTS.
The information being deleted is still required to be performed and
is being performed by the licensee because the information is
contained in regulatory requirements contained in the Code of
Federal Regulations. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change will not impose any different
operational requirements. Thus, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change will not reduce a margin of safety because
it has no impact on
[[Page 66472]]
any safety analysis assumptions. This change is administrative in
nature. The requirements being deleted from the CTS are still
required to be met and are being met by the licensee because these
requirements exist in the Code of Federal Regulations. As such, no
question of safety is involved.
Items Deleted from the Technical Specifications That Have No
Application in the Proposed HBPP PDTS
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change involves deleting information from the CTS.
The deletion process involves no technical changes to the CTS. As
such, this change is administrative in nature and does not impact
initiators of analyzed events or assumed mitigation of accidents or
transient events. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change will not impose any different
operational requirements. Thus, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change will not reduce a margin of safety because
it has no impact on any safety analysis assumptions. This change is
administrative in nature. As such, no question of safety is
involved.
Relocating Information from CTS to the DSAR, PDTS Bases or Other
Licensee-Controlled Documents
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change relocates requirements and descriptive
information from the CTS to the PDTS Bases, DSAR, or other licensee-
controlled documents. The PDTS Bases, DSAR, or other licensee-
controlled documents containing the relocated requirements and
information will be maintained using provisions of 10 CFR 50.59 or
other appropriate regulatory controls. Since any future changes to
the PDTS Bases, DSAR, or other licensee-controlled documents will be
evaluated per the requirements of 10 CFR 50.59 or other appropriate
regulatory controls, proper controls are in place to adequately
limit the probability or consequences of an accident previously
evaluated. Therefore, this change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change will not impose any different
requirements and adequate control of the information will be
maintained. Thus, this change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change will not reduce a margin of safety because
it has no impact on any safety analysis assumptions. In addition,
the requirements and information to be relocated from the CTS to the
PDTS Bases, DSAR, or other licensee-controlled documents are not
being revised; they are being relocated verbatim. Since any future
changes to these requirements in the PDTS Bases, DSAR, or other
licensee-controlled documents will be evaluated per the requirements
of 10 CFR 50.59 or other appropriate regulatory controls, proper
controls are in place to maintain an appropriate margin of safety.
Therefore this change does not involve a significant reduction in a
margin of safety.
Addition of New Requirements or New Sections to the PDTS
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change involves the addition of requirements or
sections to the proposed PDTS. Each addition either provides
equivalent or potentially more restrictive controls than previously
provided. The additional requirements or controls do not impact
initiators of analyzed events or assumed mitigation of accidents or
transient events. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change will not impose any different
operational requirements and any addition is non-operational in
nature and has been identified and justified. Thus, this change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change will not reduce a margin of safety because
it has no impact on any safety analysis assumptions. This change
provides the equivalent or more restrictive requirements on the
surveillance and control of TS parameters. As such, no question of
safety is involved.
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: Christopher J. Warner, Esquire, Pacific Gas
and Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: August 22, 2001.
Description of amendment request: Approve reactor core power
uprate, and revise the Technical Specifications to reflect the power
uprate.
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:
STPNOC [South Texas Project Nuclear Operating Company] has
evaluated whether or not a significant hazards consideration is
involved with the proposed amendment by focusing on the three
standards set forth in 10 CFR 50.92, ``Issuance of amendment,'' as
discussed below.
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The comprehensive analytical efforts performed to support the
proposed uprate conditions include a review and evaluation of all
components and systems (including interface systems and control
systems) that could be affected by this change. The revised power
uprate value was input to applicable safety analyses. The proposed
change is not an initiator of any design-basis accident. All of the
Nuclear Steam Supply System or Balance of Plant interface systems
will continue to perform their intended design functions and meet
all performance requirements. The primary loop components (reactor
vessel, reactor internals, control rod drive mechanisms, loop piping
and supports, reactor coolant pump, steam generator, and
pressurizer) continue to comply with their applicable structural
limits and will continue to perform their intended design functions.
Therefore, there is no increase in the probability of a structural
failure of these components.
The auxiliary systems and components continue to comply with
applicable structural limits and will continue to perform their
intended design functions. Therefore, there is no increase in the
probability of a structural failure of these components. The steam
generator safety valves will provide adequate relief capacity to
maintain the steam generators within design limits. The steam dump
system will still relieve 40
[[Page 66473]]
percent of the maximum full-load steam flow.
Therefore, the proposed change does not involve a significant
increase in the probability of an accident previously evaluated.
The applicable analyses have been evaluated with respect to the
increase in core power associated with this change. All applicable
radiological acceptance criteria continue to be met. Therefore, the
proposed change does not involve a significant increase in the
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change neither causes the initiation of any
accident nor creates any new limiting single failures. All of the
affected systems and components continue to perform their intended
design functions. The proposed change has no adverse effects on any
safety-related system or component and does not challenge the
performance or integrity of any safety-related system.
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 WRB-2M DNB methodology is used to demonstrate that core
thermal-hydraulic limits are maintained without any significant
reduction in margin of safety for the uprated power level of 3853
MWt (1.4-percent uprate) assuming core designs composed of Robust
Fuel Assemblies. The WRB-1 DNB correlation demonstrates that there
is no significant reduction in the margin of safety for core designs
composed of standard or Vantage 5 Hybrid (V5H) fuel types. Extensive
analyses of the primary fission product barriers have concluded that
all relevant design criteria remain satisfied, both from the
standpoint of the integrity of the primary fission product barrier
and from the standpoint of compliance with the regulatory acceptance
criteria. As appropriate, all evaluations have been performed using
methods that either have been reviewed and approved by the Nuclear
Regulatory Commission or are in compliance with all applicable
regulatory review guidance and standards.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, STPNOC concludes that the proposed amendment
presents no significant hazards consideration under the standards
set forth in 10 CFR 50.92(c), and accordingly, a finding of ``no
significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Robert A. Gramm.
TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: October 23, 2001.
Brief description of amendments: The proposed amendment would
revise Technical Specification (TS) 5.5.9, ``Steam Generator Tube
Surveillance Program,'' to permit tube sleeving repair techniques,
developed by Westinghouse Electric Company (Westinghouse) and referred
to as ``Westinghouse Leak Tight Sleeves.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The Westinghouse Leak Tight Sleeves are designed using the
applicable American Society of Mechanical Engineers (ASME) Boiler
and Pressure Vessel Code and, therefore, meet the design objectives
of the original steam generator tubing. The applicable design
criteria for the sleeves conforms to the stress limits and margins
of safety of Section III of the ASME code. Mechanical testing has
shown that the structural strength of repair sleeves under normal,
upset, and faulted conditions provides margin to the acceptance
limits. These acceptance limits bound the most limiting (three times
normal operating pressure differential) burst margin recommended by
Draft Regulatory Guide 1.121. Burst testing of sleeved tubes has
demonstrated that no unacceptable levels of primary-to-secondary
leakage are expected during any plant condition.
Evaluation of the repaired steam generator tubes indicates no
detrimental effects on the sleeve or sleeve-tube assembly from
reactor coolant system flow, primary or secondary coolant
chemistries, thermal conditions or transients, or pressure
conditions as may be experienced at CPSES [Comanche Peak Steam
Electric Station]. Corrosion testing of sleeve-tube assemblies
indicates no evidence of sleeve or tube corrosion considered
detrimental under anticipated service conditions.
The installation of the proposed sleeves is controlled via the
sleeving vendor's proprietary processes and equipment. The
Westinghouse process has been in use since 1984 and has been
implemented more than 24 times for the installation of over 4,200
sleeves. The CPSES steam generator design was reviewed and found to
be compatible with the installation processes and equipment.
The implementation of the proposed amendment has no significant
effect on either the configuration of the plant or the manner in
which it is operated. The consequences of a hypothetical failure of
the sleeved tube is bounded by the current steam generator tube
rupture (SGTR) analysis described in the CPSES FSAR [Final Safety
Analysis Report]. Due to the slight reduction in diameter caused by
the sleeve wall thickness, primary coolant release rates would be
slightly less than assumed for the steam generator tube rupture
analysis, depending on the break location, and therefore, would
result in lower total primary fluid mass release to the secondary
system. A main steam line break or feed line break will not cause a
SGTR since the sleeves are analyzed for a maximum accident
differential pressure greater than that predicted in the CPSES
safety analysis. The proposed reduction of the steam generator
primary to secondary operational leakage limit provides added
assurance that leaking flaws will not propagate to burst prior to
commencement of plant shutdown.
In conclusion, based on the discussion above, these changes will
not significantly increase the probability or consequences of an
accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The Westinghouse Leak Tight Sleeves are designed using the
applicable ASME Code as guidance; therefore, they meet the
objectives of the original steam generator tubing. As a result, the
functions of the steam generators will not be significantly affected
by the installation of the proposed sleeves. The proposed repair
sleeves do not interact with any other plant systems. Any accident
as a result of potential tube or sleeve degradation in the repaired
portion of the tube is bounded by the existing tube rupture accident
analysis. The continued integrity of the installed sleeve is
periodically verified by the Technical Specification requirements.
The implementation of the proposed amendment has no significant
effect on either the configuration of the plant or the manner in
which it is operated. As discussed above, the reduced primary to
secondary leakage limit is considered a conservative change in the
plant limiting conditions for operation. Therefore, TXU Electric
concludes that this proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The repair of degraded steam generator tubes with Westinghouse
Leak Tight Sleeves restores the structural integrity of the degraded
tube under normal operating and postulated accident conditions. The
design safety factors utilized for the repair sleeves are consistent
with the safety factors in the ASME Code used in the original steam
generator design. The portions of the installed sleeve assembly that
represents the reactor coolant pressure boundary can be monitored
for the initiation and progression of sleeve/tube wall degradation.
Use of the
[[Page 66474]]
previously identified design criteria and design verification
testing assures that the margin of safety is not significantly
different from the original steam generator tubes. The proposed
sleeve inspection requirements are more stringent than existing
requirements for inspection of the steam generator tubes, and the
reduction in the operational limit for primary to secondary leakage
through the steam generator tubes is more conservative than current
requirements. Therefore, TXU Electric concludes that the proposed
change does not involve a significant reduction in a margin of
safety.
EPRI [Electric Power Research Institute] qualified eddy current
techniques will be used for the detection of tube degradation in 3/4
inch welded sleeved tubes. Alternate inspection techniques, may be
used as they become available, as long as it can be demonstrated
that the technique used provides the same degree or greater degree
of inspection rigor.
The effect of sleeving on the design transients and accident
analyses were reviewed and found to remain valid up to the level of
steam generator tube plugging consistent with the minimum reactor
flow rate as specified in Technical Specification 3.4.1. Continued
compliance with the RCS [Reactor Coolant System] flow limits of
Technical Specification 3.4.1 is assured through precision flow
measurements.
Because all relevant safety analyses were reviewed and found to
remain valid, and because the appropriate design margins are
maintained through compliance with the relevant ASME Code
requirements, it is concluded that 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: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: November 20, 2001.
Description of amendment request: The proposed amendment would: (1)
Move Table 4.7.2, ``Primary Containment Isolation Valves'' and
references to the Table from the Vermont Yankee Nuclear Power Station
(VY) Technical Specifications (TSs) to the Technical Requirements
Manual; (2) change Surveillance Requirement 4.7.B.1.b to reflect that
the Standby Gas Treatment system (SBGT) duct heater needs to meet
relative humidity design basis; (3) add section 3.7.E, ``Reactor
Building Automatic Ventilation System Isolation Valves,'' to the Table
of Contents; (4) remove wording in 3.5.A.4.a and b referencing a one-
time 30-day Limiting Condition for Operation; and (5) make
administrative changes to Sections 5.3 and 6.4.
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 operation of the Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
The proposed changes consist of removal of the primary containment
isolation valve component list from the VY TS, revision of the SBGT
inlet heater surveillance minimum power rating and other administrative
changes. The probability of occurrence of a previously evaluated
accident is not increased because neither containment isolation nor the
SBGT heater are accident initiators, and the proposed changes do not
impact any accident initiating conditions. The consequences of an
accident previously evaluated are not increased because the proposed
changes do not impact the ability of containment to restrict, or SBGT
to filter, the release of any fission product radioactivity to the
environment. The proposed changes to remove the primary containment
isolation valve component list from TS, relocate the information to a
licensee controlled document, and to change the SBGT inlet heater power
input surveillance requirement, will have no significant impact on any
safety related structures, systems or components. The TS requirements
for the primary containment isolation valves and SBGT operability and
surveillance will not be changed. Additionally, the administrative
changes do not affect any system operation or function.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve any physical alteration of
plant equipment and do not change the method by which any safety-
related system performs its function. No new or different types of
equipment will be installed. The proposed changes do not create any new
accident initiators or involve an activity that could be an initiator
of an accident of a different type.
Therefore, the proposed changes will not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a significant
reduction in a margin of safety.
The proposed administrative changes, the removal of the primary
containment isolation valve component list from TS and the change to
the SBGT inlet heater power input surveillance requirement, do not
alter the TS requirements for containment integrity, containment
isolation, SBGT operability, or adversely affect their capability. The
changes will not alter the basic operation of process variables,
systems, or components as described in the safety analysis. No new
equipment is introduced.
The proposed changes do no impact design margins of the primary
containment isolation system, SBGT or any other system to perform their
safety functions. The essential safety functions of providing primary
containment integrity and providing filtration of airborne radioactive
releases, are maintained. There is no physical or operational change
being made which would alter the sequence of events, plant response, or
margins in existing safety analyses. The proposed changes result in no
impact on analyzed accident event precursors or effects.
These proposed changes do not alter the physical design of the
plant. There is no change in methods of operation. The proposed changes
do not alter the means by which primary containment isolation
capability is maintained and SBGT is operated.
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: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
[[Page 66475]]
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].
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant, Middlesex County, Connecticut
Date of amendment request: May 30, 2001, as supplemented on
November 6, 2001.
Brief description of amendment: Connecticut Yankee Atomic Power
Company (the licensee) requested changes to the Technical
Specifications (TSs) for the Haddam Neck Plant. The changes to Sections
5 and 6 of the TSs correct terminology, clarify the specifications for
consistency with established programs and Standard TSs, and reflect
current plant conditions. The changes also reflect the licensee's
current organization titles. For information only, the licensee also
included proposed changes to the TS Bases for spent fuel pool water
level and cooling. The NRC staff did not review the proposed changes to
the TS Bases.
Date of issuance: December 4, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 196.
Facility Operating License No. DPR-61: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 22, 2001 (66 FR
44164).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 4, 2001.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut Inc., et al., Docket Nos. 50-336 and 50-
423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London
County, Connecticut Date of application for amendment: August 9, 2001
Brief Description of amendment: The amendments modify the Millstone
Nuclear Power Station, Unit Nos. 2 and 3 Technical Specifications to
clarify the licensed operator qualification standards.
Date of issuance: December 5, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 258 and 199.
Facility Operating License Nos. NPD-69 and NPF-49: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 17, 2001 (66 FR
52798).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 5, 2001.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: April 23, 2001.
Brief description of amendment: The amendment approves a change to
Technical Specification (TS) 3.8.1.1, ``Electrical Power System--A.C.
Sources.'' The change removes Surveillance Requirement 4.8.1.1.2.c.1
regarding Emergency Diesel Generator inspection at least once per 18
months during shutdown condition.
Date of issuance: December 7, 2001.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 259.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 12, 2001 (66 FR
31705).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 7, 2001.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina Date of
application for amendments: June 13, 2000, as supplemented August 30
and September 10, 2001.
Brief description of amendments: The amendments revised the
Facility Operating License of each unit to (1) delete license
conditions that have been fulfilled; and (2) make other corrections and
editorial changes.
Date of issuance: December 5, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 200 and 181.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Facility Operating License.
Date of initial notice in Federal Register: November 1, 2000 (65 FR
65341).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 5, 2001.
No significant hazards consideration comments received: No.
[[Page 66476]]
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: June 15, 2001.
Brief description of amendments: Eliminate the Technical
Specifications (TS) requirement that the Automatic Depressurization
System (ADS) designated Safety/Relief Valves (S/RVs) open during the
manual actuation of the ADS and rewords the Surveillance Requirement
(SR) 3.5.1.8 frequency to require the testing of all required ADS
manual actuation solenoids during the performance of SR 3.5.1.8 in
place of testing on a staggered basis.
Date of issuance: December 13, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 151 and 137.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 8, 2001 (66 FR
41618).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 13, 2001.
No significant hazards consideration comments received: No.
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 application for amendments: August 13, 2001.
Brief description of amendments: These amendments delete Technical
Specifications (TS) Section 6.8.4, which required a Post-Accident
monitoring program, for Beaver Valley Power Station, Unit Nos. 1 and 2,
and thereby eliminate the requirements to have and maintain the post-
accident sampling system (PASS) for those units.
Date of Issuance: December 6, 2001.
Effective date: Upon issuance and shall be implemented within 180
days.
Amendment Nos.: 245, 123.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 19, 2001 (66
FR 48286).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 6, 2001.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: May 17, 2001, as supplemented
by letter dated September 5, 2001.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3/4.9.3, ``Decay Time,'' to allow the start of core
offload at 100 hours after reactor subcriticality between September 15
and June 15, when the lake temperature is assumed to be not higher than
77.8 deg.F, and 148 hours after reactor subcriticality between June 16
and September 14, when the lake temperature is assumed to be not higher
than 85 deg.F. TS 3/4.9.3 currently prohibits fuel movement in the
reactor pressure vessel until the reactor has been subcritical for at
least 168 hours. The 168-hour decay time was placed in the CNP TS with
Amendment Nos. 169 and 152 to DPR-58 and DPR-74, respectively, on
January 14, 1993.
Date of issuance: November 30, 2001.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 260 and 243.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 22, 2001 (66 FR
44174)
The supplemental letter contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 30, 2001.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: August 30, 2001, as supplemented
October 10 and November 16, 2001.
Brief description of amendment: The amendment revises the Technical
Specification safety limit minimum critical power ratio for two
recirculation pump operation for Cycle 21.
Date of issuance: December 6, 2001.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 125.
Facility Operating License No. DPR-22: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 3, 2001 (66 FR
50470)
The October 10 and November 16, 2001, supplements provided
clarifying information that was within the scope of the original
Federal Register notice and did not change the staff's initial proposed
no significant hazards considerations determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated December 6, 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: February 7, 2001, as supplemented by
letters dated October 17 and November 2, 2001.
Brief description of amendment: The requested changes replaced the
current accident source term used in the design basis radiological
analyses for control room habitability with an alternative source term
(AST) pursuant to 10 CFR 50.67, ``Accident Source Term.'' OPPD
requested a full implementation of the AST. Changes were also made to
the Ft. Calhoun Technical Specifications to make them consistent with
the revised associated accident analysis.
Date of issuance: December 5, 2001.
Effective date: December 5, 2001, to be implemented within 60 days
from the date of issuance.
Amendment No.: 201.
Facility Operating License No. DPR-40: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 2, 2001 (66 FR
22031).
The October 17 and November 2, 2001, supplemental letters provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 5, 2001.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: May 3, 2001.
Brief Description of amendments: The amendments relocate cycle-
specific
[[Page 66477]]
reactor coolant system parameter limits from the Technical
Specifications (TS) and associated Bases, to the Core Operating Limits
Report. The amendments also, add a reference to the Refueling Boron
Concentration to TS 5.6.5 to correct an omission.
Date of issuance: December 4, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 151 and 143.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: October 31, 2001 (66 FR
55024).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 4, 2001.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: February 12, 2001.
Brief description of amendments: The amendments consist of deleting
Surveillance Requirement 4.4.6.2.2.e of South Texas Project Technical
Specifications Section 3/4.4.6.2.
Date of issuance: December 11, 2001.
Effective date: As of the date of issuance, and shall be
implemented within 30 days from the date of issuance.
Amendment Nos.: Unit 1--134; Unit 2--123.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 12, 2001 (66 FR
31715).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 11, 2001.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
Date of application for amendment: December 14, 2000.
Brief description of amendment: These amendments revise Technical
Specifications Sections 4.7.7.1.d.1 and 4.7.7.2.a. These changes
increase the specified minimum number of compressed bottles of air from
84 to 102, and revise the differential pressure limit across the
Control Room Emergency Ventilation System HEPA Filter, demister filter,
and charcoal adsorber.
Date of issuance: December 12, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 228 and 209.
Facility Operating License Nos. NPF-4 and NPF-7: Amendments change
the Technical Specifications.
Date of initial notice in Federal Register: January 24, 2001 (66 FR
7687).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 12, 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 17th day of December, 2001.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Acting Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 01-31473 Filed 12-21-01; 8:45 am]
BILLING CODE 7590-01-P