[Federal Register Volume 66, Number 124 (Wednesday, June 27, 2001)]
[Notices]
[Pages 34278-34293]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-15818]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from June 4 through June 15, 2001. The last
biweekly notice was published on June 12, 2001 (66 FR 31700).
Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received
[[Page 34279]]
within 30 days after the date of publication of this notice will be
considered in making any final determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC Public Document
Room, located at One White Flint North, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. The filing of requests for a hearing
and petitions for leave to intervene is discussed below.
By July 27, 2001, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, located at One
White Flint North, 11555 Rockville Pike (first floor), Rockville,
Maryland 20852. Publicly available records will be accessible and
electronically from the ADAMS Public Library component on the NRC Web
site, http://www.nrc.gov (the Electronic Reading Room). If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Branch, or may be delivered to the Commission's Public
Document Room, located at One White Flint North, 11555 Rockville Pike
(first floor), Rockville, Maryland 20852, by the above date. A copy of
the petition should also be sent to the Office of the General Counsel,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to
the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be 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
[[Page 34280]]
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].
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: June 4, 2001.
Description of amendments request: The proposed license amendments
revise, from 2 hours to 6 hours, the time period in Surveillance
Requirement 3.6.1.6.1 for verifying that each suppression chamber-to-
drywell vacuum breaker is closed after any discharge of steam to the
suppression chamber from any source. In conjunction with this change,
the Completion Time associated with Required Action B.1 for closing an
open vacuum breaker is being revised from 8 hours to 4 hours.
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 license amendments do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes provide additional time to verify that each
vacuum breaker is closed and reduce the time allowed for closing an
open vacuum breaker. The safety functions of the suppression
chamber-to-drywell vacuum breaker valves are to relieve vacuum in
the drywell following a postulated loss-of-coolant accident and to
remain closed, except when the vacuum breakers are performing their
intended design function, in order to ensure that no excessive
bypass leakage occurs from the drywell to the suppression chamber.
With a vacuum breaker not closed, communication between the drywell
and suppression chamber airspaces could occur and, if a loss-of-
coolant accident were to occur, there would be the potential for
primary containment overpressurization due [to] steam leakage from
the drywell to the suppression chamber without quenching. The vacuum
breakers do not perform a safety function that initiates, or alters
initiation of, an accident previously evaluated. Rather, the vacuum
breakers function to mitigate the consequences of certain design
basis accidents. Therefore, the proposed changes do not involve an
increase in the probability of an accident previously evaluated or
the method of performing their safety functions.
As noted above, the vacuum breakers function to mitigate the
consequences of certain design basis accidents. The proposed changes
to the Surveillance Requirement and Completion Time provide
additional time to verify that each vacuum breaker is closed and
reduce the time allowed for closing an open vacuum breaker; however,
the proposed changes do not alter the safety functions of the vacuum
breakers. When performing the surveillance to verify each vacuum
breaker is closed, the expected result is the verification that the
component is indeed closed. However, if this surveillance result is
not obtained, the Technical Specifications limit the time allowed to
close the vacuum breaker. Additional time is being provided to
verify that each vacuum breaker is closed; however, the overall time
allowed for closing and verifying closure of a vacuum breaker is not
being increased. Since the overall time to take action for an open
vacuum breaker has not been increased, the proposed changes do not
involve an increase in the consequences of an accident previously
evaluated.
2. The proposed license amendments will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The suppression chamber-to-drywell vacuum breakers are not an
initiator of any design basis accident. Rather, the safety functions
of the vacuum breaker valves are to relieve vacuum in the drywell
following a loss-of-coolant accident and to remain closed when not
relieving vacuum to ensure that no excessive bypass leakage occurs
from the drywell to the suppression chamber. Neither safety function
of these vacuum breakers is altered by the proposed changes.
Therefore, the proposed changes will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed license amendments do not involve a significant
reduction in a margin of safety.
The proposed changes will not affect the ability of the
suppression chamber-to-drywell vacuum breakers to perform their
safety functions. Rather, as previously stated, the proposed changes
provide additional time to verify that each vacuum breaker is closed
and reduce the time allowed for closing an open or inoperable vacuum
breaker. As a result, the overall time for taking action for an open
vacuum breaker is unchanged. The vacuum breakers will continue to be
verified closed every 14 days, as part of a required functional test
of the vacuum breaker every 31 days, and following any activity
involving the discharge of steam to the suppression chamber. If a
vacuum breaker is found to be open and cannot be closed as required,
plant shutdown will continue to be required within the same time
requirements as currently specified in the Technical Specifications.
Current Technical Specifications allow up to 10 hours to close an
open vacuum breaker (i.e., 2 hours to perform the surveillance to
verify vacuum breaker closure and, if necessary, 8 hours to close
the vacuum breaker). The proposed change maintains the 10 hour limit
by reducing the time to 4 hours to close an open or inoperable
vacuum breaker while increasing the time to 6 hours to complete the
surveillance to verify vacuum breaker closure. Thus, on this basis,
the proposed license amendments will not change overall plant risk
and do not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Patrick M. Madden, Acting.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: May 18, 2001.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3/4.9.4 ``Containment Building
Penetrations'' and the associated Bases to permit containment building
penetrations to remain open, under administrative controls, during core
alterations or the movement of irradiated fuel within the containment.
Specifically, the licensee proposes: (1) Incorporating an alternate
source term methodology in the fuel handling accident analysis; (2)
revising TS 3.9.4 to remove portions of a note restricting the
applicability of administrative controls with respect to containment
penetrations; and (3) including the use of administrative controls on
the equipment hatch and other penetrations that provide access from
containment atmosphere to outside atmosphere.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes modify TS requirements previously reviewed
and
[[Page 34281]]
approved by the NRC in improved Technical Specifications (ITS) and
changes to ITS as described in TSTF [Technical Specification Task
Force]-312. An alternate source term calculation has been performed
for the HNP [Harris Nuclear Plant] that demonstrates that dose
consequences remain below limits specified in NRC Regulatory Guide
1.183 and 10 CFR 50.67. The proposed change does not modify the
design or operation of equipment used to move spent fuel or to
perform core alterations[.]
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Containment penetrations are designed to form part of the
containment pressure boundary. The proposed change provides for
administrative controls and operating restrictions for containment
penetrations consistent with guidance approved by the NRC staff.
Containment penetrations are not an accident initiating system as
described in the Final Safety Analysis Report [FSAR]. The proposed
change does not affect other Structures, Systems, or Components. The
operation and design of containment penetrations in operational
modes 1-4 will not be affected by this proposed change.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed changes modify required Actions and Surveillance
Requirements previously reviewed and approved by the NRC in improved
Technical Specifications (ITS) and changes to ITS, TSTF-312.
Additionally, the implementation of the alternate source term
methodology is consistent with NRC Regulatory Guide 1.183. The
proposed change to containment penetrations does not significantly
affect any of the parameters that relate to the margin of safety as
described in the Bases of the TS or the FSAR. Accordingly, NRC
Acceptance Limits are not significantly affected by this change.
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: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Patrick M. Madden, Acting.
Duke Energy Corporation, et al., Docket No. 50-414, Catawba Nuclear
Station, Unit 2, York County, South Carolina
Date of amendment request: March 9, 2001.
Description of amendment request: The amendment will revise the
cold leg elbow tap flow coefficients used in the determination of
Reactor Coolant System (RCS) flow rate at Catawba Nuclear Station, Unit
2. No changes in Technical Specification are necessary for this
amendment.
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. No component modification, system realignment, or change
in operating procedure will occur which could affect the probability
of any accident or transient. The revised cold leg elbow tap flow
coefficients will not change the probability of actuation of any
Engineered Safeguards Feature or other device. The actual Unit 2 RCS
flow rate will not change. Therefore, the consequences of previously
analyzed accidents will not change as a result of the revised flow
coefficients.
Second Standard
The proposed amendment will not create the possibility of a new
or different kind of accident from any accident previously
evaluated. No component modification or system realignment will
occur which could create the possibility of a new event not
previously considered. No change to any methods of plant operation
will be required. The elbow taps are already in place, and are
presently being used to monitor flow for Reactor Protection System
purposes. They will not initiate any new events.
Third Standard
The proposed amendment will not involve a significant reduction
in a margin of safety. The removal of some of the excess flow
margin, which was introduced by the hot leg streaming flow penalties
in later calorimetrics, will allow additional operating margin
between the indicated flow and the Technical Specification minimum
measured flow limit. The proposed changes in the cold leg elbow tap
flow coefficients will continue to be conservative with respect to
the analytical model flow predictions, since the proposed
coefficients will continue to contain some hot leg streaming
penalties from the calorimetric determined coefficients used in the
average.
An increase in the RCS flow indication of approximately 1.0%
will increase the margin to a reactor trip on low flow but will not
adversely affect the plant response to low flow transients. Current
UFSAR Chapter 15 transients that would be expected to cause a
reactor trip on the RCS low flow trip setpoint are Partial Loss of
Reactor Coolant Flow, Reactor Coolant Pump Shaft Seizure and Reactor
Coolant Pump Shaft break transients. Three reactor trip functions
provide protection for these transients, RCS low flow reactor trip,
RCP undervoltage reactor trip and RCP underfrequency reactor trip.
The transient analyses of these events assume the reactor is tripped
on the low flow reactor trip setpoint. This is conservative and
produces a more severe transient response since a reactor trip on
undervoltage or underfrequency would normally be expected to trip
the reactor sooner and therefore reduce the severity of these
transients.
The RCS low flow reactor trip is currently set at 91% of the
Technical Specification minimum measured flow of 390,000 gpm. The
setpoint will not be revised as a result of this change, which means
the transients relying on this function will behave in the same
manner with the reactor trips occurring at essentially the same
conditions as previously analyzed. Therefore, any small increase in
the reactor trip margin gained by the small increase in the
indicated RCS flow will not adversely affect the plant response
during these low flow events.
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: Ms. Lisa F. Vaughn , Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: Richard L. Emch, Jr.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: May 23, 2001.
[[Page 34282]]
Description of amendment request: The amendment request proposes a
change to the minimum critical power ratio safety limit (SLMCPR) and
changes to the references for the analytical methods used to determine
the core operating 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. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
The Minimum Critical Power Ratio (MCPR) safety limit is defined
in the Bases to Technical Specification [TS] 2.1.1 as that limit
which ``ensures that during normal operation and during AOOs
[Anticipated Operational Occurrences], at least 99.9% of the fuel
rods in the core do not experience transition boiling.'' The MCPR
safety limit satisfies the requirements of General Design Criterion
10 of Appendix A to 10 CFR [Part] 50 regarding acceptable fuel
design limits. The MCPR safety limit is re-evaluated for each reload
using NRC [Nuclear Regulatory Commission]-approved methodologies.
The analyses for RBS [River Bend Station] Cycle 11 have concluded
that a two-loop MCPR safety limit of 1.08, based on the application
of Framatome ANP Richland, Inc.'s [FRA-ANP] [(proprietary)] NRC-
approved MCPR safety limit methodology, will ensure that this
acceptance criterion is met. For single-loop operation, a MCPR
safety limit of 1.10, also ensures that this acceptance criterion is
met.
In addition to the MCPR safety limit, core operating limits are
established to support the Technical Specification 3.2 requirements
which ensure that the fuel design limits are not exceeded during any
conditions of normal operation or in the event of any anticipated
operational occurences (AOO). The methods used to determine the core
operating limits for each operating cycle are based on methods
previously found acceptable by the NRC and listed in TS section
5.6.5. A change to TS section 5.6.5 is requested to include the FRA-
ANP methods in the list of NRC approved methods applicable to RBS.
These NRC approved methods will continue to ensure that acceptable
operating limits are established to protect the fuel cladding
integrity during normal operation and in the event of an AOO.
The requested Technical Specification changes do not involve any
plant modifications or operational changes that could affect system
reliability or performance or that could affect the probability of
operator error. The requested changes do not affect any postulated
accident precursors, do not affect any accident mitigating systems,
and do not introduce any new accident initiation mechanisms.
Therefore, these changes to the Minimum Critical Power Ration
(MCPR) safety limit and to the list of methods used to determine the
core operating limits do not involve a significant increase in the
probability or consequences of any accident previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
The ATRIUM-10 fuel to be used in Cycle 11 is of a design
compatible with the co-resident GE-11. Therefore, the introduction
of ATRIUM-10 fuel into the Cycle 11 core will not create the
possibility of a new or different kind of accident. The proposed
changes do not involve any new modes of operation, any changes to
setpoints, or any plant modifications. The proposed revised MCPR
safety limits have accounted for the mixed fuel core and have been
shown to be acceptable for Cycle 11 operation. Compliance with the
criterion for incipient boiling transition continues to be ensured.
The core operating limits will continue to be developed using NRC
approved methods which also account for the mixed fuel core design.
The proposed MCPR safety limits or methods for establishing the core
operating limits do not result in the creation of any new precursors
to an accident.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
The MCPR safety limits have been evaluated in accordance with
Framatome ANP Richland, Inc.'s NRC-approved cycle-specific safety
limit methodology to ensure that during normal operation and during
Anticipated Operational Occurrences (AOO's) at least 99.9% of the
fuel rods in the core are not expected to experience transition
boiling. On this basis, the implementation of this Framatome ANP
Richland, Inc. methodology does not involve a significant reduction
in a margin of safety.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: May 11, 2001.
Description of amendment request: This amendment revised the
Technical Specifications to allow, on a one-time basis only, Entergy
Nuclear Operations, Inc. to extend the allowed out-of-service time for
the Residual Heat Removal Service Water (RHRSW) System from 7 days to
11 days. This amendment is only applicable during installation of the
modification 00-12 to the ``B'' RHRSW Strainer.
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:
Involve an increase in the probability or consequences of an
accident previously evaluated.
The CCDP [Conditional Core Damage Probability] due to this
proposed change is calculated to be 4.33 E-8 (assuming no-risk
significant SSC maintenance), which falls below the threshold
probability of 1 E-6 for risk significance of temporary changes to
the plant configuration in the EPRI PSA Applications Guide
(Reference 2). The ICLERP [incremental conditional large early
release probability] is calculated to be 8.85 E-8, which falls below
the threshold probability of 1 E-7 for risk significance per
Reference 2 [see application dated May 11, 2001].
This proposed change does not increase the consequences of an
accident previously evaluated because all relevant accidents (LOCA)
[loss-of-coolant accident] would result in the transfer of decay
heat to the suppression pool. For this scenario, the same compliment
of equipment will be available to achieve and maintain cold shutdown
as is required by the current TS LCO [limiting condition for
operation].
Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change does not physically alter the plant. As
such, no new or different types of equipment will be installed. The
new design for the RHRSW strainer packing gland will be evaluated
under a separate 10 CFR 50.59 evaluation and is considered to be
functionally equivalent for the purposes of this one-time-only
proposed TS change.
The connection and use of a temporary hose for achieving limited
containment heat removal in the event the ``A'' division of RHRSW is
rendered inoperable for some reason is a contingency plan that is
already addressed by current plant procedures.
Involve a significant reduction in a margin of safety.
The CCDP due to this proposed change is calculated to be 4.33 E-
8 (assuming no-risk significant SSC maintenance). This value falls
below the threshold probability of 1 E-6 for risk significance of
temporary changes to the plant configuration in the EPRI PSA
Applications Guide (Reference 2). The CLERP is calculated to be 8.85
E-8, which falls below the threshold probability of 1 E-7 for risk
significance per Reference 2.
The consequences of a postulated accident occurring during the
extended allowable out-
[[Page 34283]]
of-service time are bounded by existing analyses, therefore, there
is no significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Section Chief: Richard P. Correia, Acting.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: May 22, 2001.
Description of amendment request: The proposed change to Technical
Specification (TS) 3/4.7.1.2, Emergency Feedwater (EFW) System expands
and clarifies the current TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
The administrative and more restrictive changes will not affect
the assumptions, design parameters, or results of any accident
previously evaluated. The accident mitigation features of the plant
are not affected by these proposed changes. The proposed changes do
not add or modify any existing equipment. The administrative change
to test EFW pumps pursuant to the Inservice Test Program will ensure
the EFW pumps are tested against the more restrictive of the data
points required by either the safety analysis or the Inservice Test
Program. Therefore, the proposed administrative changes do not
involve a significant increase in the probability or consequences of
any accident previously evaluated.
The less restrictive changes (allowing 7 days for an inoperable
pump due to an inoperable steam supply, allowing 24 hours for an
inoperable steam supply and one inoperable motor driven EFW pump,
allowing 72 hours for two inoperable motor driven EFW pumps,
performing Surveillance Requirements during other than shutdown
conditions, allowing the use of actual actuation signals in addition
to test signals, and delaying the requirement to complete
Surveillance Requirement ``d'' to just prior to Mode 2) will not
affect the assumptions, design parameters, or results of any
accident previously evaluated. The accident mitigation features of
the plant are not affected by these proposed changes. The proposed
changes do not add or modify any existing equipment. Therefore, the
proposed less restrictive changes do not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
The proposed changes do not alter the design or configuration of
the plant. There has been no physical change to plant systems,
structures, or components. The proposed changes will not reduce the
ability of any of the safety-related equipment required to mitigate
Anticipated Operational Occurrences or accidents.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
The proposed change to the LCO [Limiting Conditions for
Operation] requiring three pumps and two flow paths be OPERABLE
maintains the functionality of the EFW such that it is capable of
performing its design function as assumed in the Final Safety
Analysis Report. If the functionality of the system is not
maintained, Technical Specifications require ACTIONs be taken,
within specified time limitations, to restore EFW to OPERABLE status
or shutdown the reactor. This action is consistent with the existing
Technical Specifications and NUREG-1432.
The allowed outage time for one inoperable steam supply has been
increased from 72 hours to 7 days in accordance with NUREG-1432.
This is acceptable due to the redundant OPERABLE steam supply, the
availability of redundant OPERABLE motor-driven EFW pumps, and the
low probability of an event requiring the inoperable steam supply.
This change is consistent with NUREG-1432 and has therefore been
previously approved by the NRC [Nuclear Regulatory Commission].
The ACTION for an inoperable steam supply to the turbine-driven
EFW pump steam turbine concurrent with one motor-driven EFW pump
being inoperable will allow a 24 hour completion time. This change
is acceptable based on the ability of the system to cool the reactor
coolant system to shutdown cooling entry conditions following a loss
of normal feedwater. The 24 hour completion time is reasonable based
on the redundant OPERABLE steam supply to the turbine-driven EFW
pump steam turbine, the OPERABLE motor-driven EFW pump, and the low
probability of an event requiring the inoperable steam supply to the
turbine-driven EFW pump.
The ACTION for an inoperable steam supply to the turbine-driven
EFW pump steam turbine concurrent with both motor-driven EFW pumps
being inoperable as proposed requires a unit shutdown be initiated
immediately. This change is appropriate due to the seriousness of
the condition and is acceptable due to the ability of the EFW system
to support the unit shut down.
The ACTION for the EFW system inoperable for reasons other than
those described in ACTION (a), (b), or (c) and able to deliver at
least 100% flow to either steam generator as proposed will allow a
72 hour completion time. This change is acceptable based on the
ability of the system to cool the RCS [Reactor Coolant System] to
SDC [Shutdown Cooling] entry conditions following a design basis
accident assuming no single active failure.
The ACTION for the EFW system inoperable for reasons other than
those described in ACTION (a), (b), or (c) and able to deliver at
least 100% combined flow to the steam generators as proposed
requires a unit shutdown be initiated immediately. This change is
appropriate due to the seriousness of the condition and is
acceptable due to the ability of the EFW system to support the unit
shut down.
The ACTION for the EFW system inoperable and unable to deliver
at least 100% flow to the steam generators as proposed requires
immediate action be taken to restore the ability to deliver at least
100% flow to the steam generators. The unit is in a seriously
degraded condition in that the EFW system is unable to support a
unit shutdown. This change is consistent with the intent of the
current EFW Technical Specification and NUREG-1432.
Testing pursuant to Specification 4.0.5 (Inservice Testing
Program) as proposed for Surveillance Requirement `b' will ensure
the EFW pumps are tested against the more restrictive of the data
points required by either the safety analysis or ASME [American
Society of Mechanical Engineers] Section XI.
The remaining changes to the EFW Technical Specification are
consistent (other than format) with NUREG-1432 and have therefore
been previously approved by the NRC.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn
1400 L Street NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Exelon Generation Company, LLC, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois.
Date of amendment request: February 28, 2001
Description of amendment request: The proposed amendments would
[[Page 34284]]
revise the Technical Specifications to eliminate the requirement for at
least one person qualified to stand watch to be present in the control
room when nuclear fuel is stored in the spent fuel pool.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications (TS) change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The Defueled Safety Analysis Report (DSAR) identifies three
categories of events: spent fuel pool events (i.e., operational
occurrences), fuel handling accidents in the fuel building, and
radioactive waste handling accidents. There are no active controls
in the control room that affect spent fuel pool equipment, or the
handling of fuel or radioactive waste. Actions to mitigate the
consequences of these events are taken outside the control room.
Emergency response is not adversely affected by this proposed change
because the control room is still available to the emergency
response team and communication capability and timeliness will not
be affected. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The configuration, operation and accident response of the
systems, structures or components that support safe storage of the
spent fuel are unchanged by the proposed TS change. Current site
surveillance requirements ensure frequent and adequate monitoring of
system and component functionality. Systems in the Spent Fuel
Nuclear Island will continue to be operated in accordance with
current design requirements and no new components or system
interactions have been identified. No new accident scenarios,
failure mechanisms or limiting single failures are introduced as a
result of the proposed change. The proposed TS change does not have
an adverse affect on any system related to safe storage of spent
fuel. Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
All design basis accident acceptance criteria will continue to
be met. The margin of safety relative to the cooling of the spent
fuel is unaffected by the proposed change as the SFP [spent fuel
pool] parameters will continue to be monitored at the same frequency
that they are monitored now. The ability of the shift crew to
respond to abnormal or accident conditions is unaffected by the
proposed change since all controls are located in the fuel building
and any necessary communication will be handled by the DERO
[Defueled Emergency Response Organization]. Therefore, it is
concluded that the proposed TS 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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Robert Helfrich, Senior Counsel,
Nuclear, Mid-West Regional Operating Group, Exelon Generation Company,
LLC, 1400 Opus Place, Suite 900, Downers Grove, Illinois 60515.
NRC Section Chief: Robert A. Gramm.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: May 15, 2001.
Description of amendment request: The proposed amendment would
revise refueling operation Technical Specification (TS) requirements
for containment equipment hatch cover closure during core alterations
and during movement of irradiated fuel both inside containment and in
the spent fuel pool or cask pit. The proposed change would allow the
containment equipment hatch cover to be off during core alterations and
movement of irradiated fuel provided the Emergency Ventilation System
is operable with the ability to filter any radioactive release. The
proposed changes involve TS 3/4.9.4, Refueling Operations -Containment
Penetrations, and TS 3/4.9.12, Refueling Operations--Storage Pool
Ventilation, and associated Bases.
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 because no such accidents are affected
by the proposed changes. The amendment application proposes to
revise DBNPS TS 3/4.9.4, Refueling Operations--Containment
Penetrations, and its associated Bases, and TS 3/4.9.12, Refueling
Operations--Storage Pool Ventilation, and its associated Bases. The
proposed changes would provide for access to the containment through
the containment equipment hatch during core alterations and movement
of irradiated fuel, provided that an Emergency Ventilation System is
operable with the ability to filter any radioactivity release
through the containment equipment hatch. The proposed changes would
also permit relying on the closing the containment personnel air
lock by a designated individual to establish the negative pressure
boundary for the Emergency Ventilation System servicing the storage
pool. The use of a designated individual to close the containment
personnel airlock is currently permitted by TS 3.9.4 for meeting
containment closure requirements. Neither the containment equipment
hatch nor the Emergency Ventilation System contributes to the
initiation of any accident described in the DBNPS Updated Safety
Analysis Report.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because no equipment, accident
conditions, or assumptions are affected which could lead to a
significant increase in radiological consequences. The approved
analysis for the fuel handling accident inside containment does not
take credit for containment closure or Emergency Ventilation System
filtering. This analysis results in a maximum calculated offsite
does well within the limits of 10 CFR 100.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because no new or
different accident initiators are introduced by these proposed means
to mitigate the consequences of an accident.
3. Not involve a significant reduction in a margin of safety
because there are no changes to the initial conditions contributing
to accident severity or the resulting consequences. Consequently,
there are no significant reductions in a margin of safety.
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.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Dade County, Florida
Date of amendment request: May 14, 2001.
[[Page 34285]]
Description of amendment request: The proposed amendments would
delete Technical Specifications (TS) Figures 5.1-1, ``Site Area Map,''
and 5.1-2, ``Plant Area Map,'' and would replace TS 5.1, ``Site,'' with
a site location description. Conforming changes are requested to delete
TS 5.1.1, ``Exclusion Area,'' TS 5.1.2, ``Low Population Zone,'' and TS
5.1.3, ``Map Defining Unrestricted Areas and Site Boundary for
Radioactive Gaseous and Liquid Effluents,'' from TS 5.1 and the TS
Index. These changes conform to NUREG-1431, Rev. 1, Improved Standard
TS for Westinghouse Plants, and the requirements of 10 CFR 50.36
(c)(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) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed amendments are administrative in nature, removing
sections and maps from the TS, which are located in other documents
previously approved by NRC. These amendments will not involve a
significant increase in the probability or consequences of an
accident previously evaluated because they do not affect assumptions
contained in plant safety analyses, the physical design and/or
operation of the plant, nor do they affect TS that preserve safety
analysis assumptions. Therefore, the proposed changes do not affect
the probability or consequences of accidents previously analyzed.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed changes to the TS are administrative in nature and
can not create the possibility of a new or different kind of
accident from any previously evaluated since the proposed amendments
will not change the physical plant or the modes of plant operation
defined in the facility operating license. No new failure mode is
introduced due to the administrative changes since the proposed
changes do not involve the addition or modification of equipment,
nor do they alter the design or operation of affected plant systems,
structures, or components.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed changes are administrative in nature and do not
affect operating limits or functional capabilities of plant systems,
structures and components. The addition of a site location
description to the TS adds geographical information to the TS.
Elimination of site and plant area maps from the TS would have no
effect on margin of safety as they are located in other controlled
plant documents. Thus, the changes proposed would not involve a
significant reduction in margin of safety of the facility.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Patrick M. Madden (Acting).
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: February 15, 2001.
Description of amendment request: The proposed amendment to the
Cooper Nuclear Station (CNS) Operating License (OL) DPR-46 would (1)
delete OL Condition 2.D, Additional Conditions for Protection of the
Environment, and (2) remove the depiction of railroad tracks in
Technical Specifications (TS) Figure 4.1-1, Site and Exclusion Area
Boundaries and Low Population Zone.
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.
OL [Operating License] Condition 2.D has become obsolete based
upon it being satisfied or superceded by amendments to the FSAR
[Final Safety Analysis Report] and OL. The previous FSAR and OL
amendments which made it obsolete were reviewed and approved based
on their individual Unreviewed Safety Question (USQ) evaluations or
no significant hazards considerations. Since this proposed change
does not physically alter any plant equipment or operating
limitations, it therefore does not impact any previously evaluated
accident initiator, nor change mitigating systems or features or
operating limitations for accidents previously evaluated in the
Updated Safety Analysis Report (USAR). Thus, it does not involve a
significant increase in the probability or consequences of an
accident previously evaluated. This is an administrative change.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This proposed change is administrative in nature. It does not
involve a physical alteration of the plant. No new or different
equipment is being installed, and no installed equipment is being
operated in a new or different manner. No setpoints for parameters
which initiate protective or mitigative action are being changed. As
a result, no new failure modes are being introduced. There are no
changes in the procedures or methods governing normal plant
operation, nor are the procedures utilized to respond to plant
transients altered as a result of this administrative change. This
change does not impose any new or different requirements or
eliminate any existing requirements. In addition, the change does
not alter assumptions made in the safety analysis, nor does it
impact the licensing basis. Therefore, the changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
Response: No.
This proposed change is administrative in nature. It does not
alter any accident analysis assumptions, conditions, or methodology.
Since this proposed change does not physically alter plant systems,
structures or components (SSC's), change mitigating systems,
features, operating limitations, nor revise accident analysis
assumptions, conditions or methodology, it does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: October 19, 2000, as supplemented March
23 and April 9, 2001.
Description of amendment request: The proposed amendment would
authorize the licensee to change the licensing basis to utilize the
full scope of an alternative radiological source term for accidents as
described in NUREG-1465, ``Accident Source Terms for Light-Water
Nuclear Power Plants,'' and change the Technical Specifications
[[Page 34286]]
to implement various assumptions in the Alternative Source Term
analyses. The portion of this amendment request regarding operability
requirements during core alterations and while moving irradiated fuel
assemblies within the secondary containment, and which provided for
selective application of the Alternative Source Term to the design-
basis fuel handling accident was previously evaluated and issued as
Amendment No. 237 on April 16, 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The Alternative Source Term and those plant systems affected by
implementing the setpoints and action levels specified in the
analyses are not assumed to initiate design basis accidents. The
Alternative Source Term does not affect the design or operation of
the facility; rather, once the occurrence of an accident has been
postulated the new source term is an input to evaluate the
consequence. The implementation of the Alternative Source Term has
been evaluated in revisions to the analyses of the limiting design
bases accidents at DAEC [Duane Arnold Energy Center]. Based on the
results of these analyses, it has been demonstrated that, with the
requested changes, the dose consequences of these limiting events
are within the regulatory guidance provided by the NRC for use with
the Alternative Source Term. This guidance is presented in NUREG
1465, 10 CFR 50.67, associated Regulatory Guide 1.183, and Standard
Review Plan (SRP) Section, 15.0.1. Since secondary containment
operability is not assumed for the fuel handling accident (FHA), the
consequences of eliminating the requirements for secondary
containment operability, secondary containment isolation valves/
dampers, secondary containment instrumentation and the Standby Gas
Treatment system during fuel movement or core alterations will not
increase the effects of a FHA beyond those evaluated in the
Alternative Source Term analysis. Therefore, the proposed changes do
not significantly increase the probability or consequences of any
previously evaluated accident.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any previously evaluated.
The Alternative Source Term and those plant systems affected by
implementing the setpoints and action levels specified in the
analyses do not initiate design basis accidents. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any previously evaluated.
3. The proposed amendment will not involve a significant
reduction in a margin of safety.
The changes proposed are associated with the implementation of a
new licensing basis for DAEC. Approval of the basis change from the
original source term developed in accordance with TID-14844 to a new
alternative source term as described in NUREG-1465 is requested by
this submittal. The results of the accident analyses revised in
support of this submittal, and the requested Technical Specification
changes, are subject to revised acceptance criteria. These analyses
have been performed using conservative methodologies. Safety margins
and analytical conservatisms have been evaluated and are satisfied.
The analyzed events have been carefully selected and margin has been
retained to ensure that the analyses adequately bound all postulated
event scenarios. The dose consequences of these limiting events are
within the acceptance criteria also found in the latest regulatory
guidance. This guidance is presented in NUREG 1465, in the approved
rulemaking for 10 CFR 50.67, and in the associated Regulatory Guide
1.183.
The proposed changes continue to ensure that the doses at the
exclusion area and low population zone boundaries, as well as the
control room, are within the corresponding regulatory limit.
Specifically, the margin of safety for these accidents is considered
to be that provided by meeting the applicable regulatory limit,
which, for most events, is conservatively set below the 10 CFR 50.67
limit. With respect to the control room personnel doses, the margin
of safety (the difference between the 10 CFR 50.67 limits and the
regulatory limit defined by 10 CFR50, Appendix A, Criterion 19 (GDC
19)) continues to be satisfied.
Therefore, because the proposed changes continue to result in
dose consequences within the applicable regulatory limits, they are
considered to not result in 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: Al Gutterman, Morgan, Lewis & Bockius, 1800
M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Claudia M. Craig.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: May 30, 2001.
Description of amendment request: The proposed amendment would
eliminate local suppression pool temperature limits from the Updated
Safety Analysis Report as the basis for limiting suppression pool
mechanical loads due to unstable steam condensation during safety
relief valve actuations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Eliminating the Local Suppression Pool Temperature Limits
(LSPTLs) will not introduce new equipment or new equipment methods
of operation, and will not alter existing system relationships.
LSPTLs are not an accident initiator and does [sic] not affect other
accident initiators. The integrity of fission product barriers do
not rely on LSPTLs since mechanical loads on containment will not be
exceeded and ECCS [emergency core cooling system] operation in the
event of an accident will not be adversely affected as demonstrated
and approved in Reference 6 [letter from G. Holahan (NRC) to R.
Pinelli (Boiling Water Reactor Owners Group), ``Transmittal of the
Safety Evaluation of General Electric Co. Topical Reports; NEDO-
30832, Entitled 'Elimination of Limit on BWR Suppression Pool
Temperature for SRV Discharge With Quenchers,' and NEDO-31695,
Entitled 'BWR Suppression Pool Temperature Technical Specification
Limits','' dated August 29, 1994].
Therefore, the proposed amendment will not significantly
increase the probability or the consequences of an accident
previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
Eliminating the LSPTLs will not introduce new equipment or new
equipment methods of operation, and will not alter existing system
relationships. Since containment integrity and ECCS operation will
not be challenged, new or different kinds of accidents are not
created.
Therefore, the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
Since LSPTLs are not required to limit mechanical loads on
containment, the margin of safety associated with containment
integrity is not significantly reduced. Since LSPTLs are not
required to prevent steam binding of the ECCS pumps, the margin of
safety associated with ECCS operation is not significantly reduced.
Therefore, the proposed amendment will not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three
[[Page 34287]]
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Section Chief: Claudia M. Craig.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1,
Washington County, Nebraska
Date of amendment request: May 15, 2001.
Description of amendment request: The proposed changes would: (1)
Replace the titles of Manager--Fort Calhoun Station and the Vice
President with generic titles, (2) relocate the requirements for the
Plant Review Committee (PRC) and the Safety Audit and Review Committee
(SARC) to the Fort Calhoun Station (FCS) Quality Assurance Program, (3)
relocate the requirements for procedure controls and records retention
to the FCS Quality Assurance Program, (4) enhance and clarify the
qualification and training requirements for individuals who perform
licensed operator functions, (5) incorporate the Westinghouse/CENP
definition of Azimuthal Power Tilt, and (6) eliminate specific mailing
address and reporting requirements that are redundant to Title 10 of
the Code of Federal Regulations (10 CFR).
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: revise the FCS definition of Azimuthal
Power Tilt, remove specific titles from the Technical
Specifications, provide minor clarifications of the training
requirements for plant staff, and indicate the change in title of
the Licensed Senior Operator. This change also relocates the
requirements for the Plant Review Committee (PRC) and the Safety
Audit and Review Committee (SARC), procedure control, and records
retention to the Fort Calhoun Station Quality Assurance Program as
described in NRC Administrative Letter 95-06.
The proposed change includes an update to the definition of
Azimuthal Power Tilt and adds the bases for the definition of
Azimuthal Power Tilt to the bases section of Section 2.10.4 as
recommended in ABB Combustion Engineering (CE) Infobulletin Number
97-07, dated December 31, 1997. As noted in the infobulletin, CE
discovered a discrepancy in the definition for CE analog plants that
use Combustion Engineering Core Operating Report (CECOR) for
monitoring and surveillance purposes. Plants that use CECOR should
use the same definition as the CE digital plants. This change will
make the FCS definition and bases agree with the improved Standard
Technical Specifications for CE digital plants, which have
previously been approved by the NRC.
The proposed change would allow the use of generic personnel
titles as provided in ANSI/ANS 3.1 and NUREG-1432, ``Standard
Technical Specifications Combustion Engineering Plants,'' in lieu of
plant-specific personnel titles. This change does not eliminate any
of the qualifications, responsibilities or requirements for these
positions, since the plant-specific personnel titles are currently
identified in licensee controlled documents such as the Updated
Safety Analysis Report (USAR) or the Quality Assurance Program. For
example, Section 12 of the Updated Safety Analysis Report describes
the management structure and reporting responsibilities of OPPD and
provides an organizational chart to determine the corporate officer
with responsibility for overall plant nuclear safety from other
corporate officers within OPPD. Therefore, changing the terminology
within the Technical Specifications, indicating this reporting
responsibility does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Changing the periodicity of review for staff overtime is also
considered an administrative change. This includes a change of the
title of the Supervisor--Operations to Manager--Shift Operations,
Licensed Senior Operator to Control Room Supervisor, and crewman to
crewmember. The change to the number of Senior Operator License
present during Core Alterations and the associated note is also
considered clarifying in nature and not a change of intent.
The proposed change would update the qualification requirements
for the Manager--Radiation Protection, the Shift Technical Advisors,
and those individuals that perform the functions described in 10 CFR
50.54(m) to Regulatory Guide 1.8, Revision 3, and ANSI/ANS 3.1-1993.
In the March 1987 revision to 10 CFR Part 55, the NRC included the
requirement that those facility licensees that have made a
commitment that is less than that required by the new rules must
conform to the new rules automatically. OPPD had previously
considered that commitments made to comply with the requirements of
NUREG-0737 and the standards applied through the Institute of
Nuclear Power Operations (INPO) accreditation process were
equivalent to the guidance provided in Regulatory Guide 1.8,
Revision 3. The proposed change provides enhancement to the current
requirements and clarifies the qualifications and training
requirements for licensed personnel. This provides additional
assurance that these personnel are properly trained and qualified
for their positions and conforms with the guidance of NRC Regulatory
Issues Summary 2001-01. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed change would relocate specific requirements for
SARC, PRC, procedure control, and records retention to the Fort
Calhoun Station Quality Assurance Program (Appendix A, of the FCS
USAR). This proposed revision does not change or eliminate
responsibilities or requirements for these programs. The management
level and expertise of personnel who are PRC or SARC members is not
being changed. The review of plant operations, procedures control,
and record retention is still required to be in compliance with the
Fort Calhoun Station Quality Assurance (QA) Program. Any changes in
the QA Program which reduce the effectiveness of the program must be
approved by the NRC in accordance with 10 CFR 50.54(a)(4). These
changes meet the criteria as described in NRC Administrative Letter
95-06. Therefore, the proposed relocation of these programs to the
QA Program does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change would also remove the requirements
prescribing specific submittal addresses, titles, and reporting
periods. For example, the requirement to submit License[e] Event
Reports within 30 days is replaced with a citation referencing 10
CFR 50.73. This is in agreement with 10 CFR 50.73 and 10 CFR
50.4(f). Additionally, an administrative requirement prescribing the
submittal of a Special Maintenance Report is being deleted, as it is
redundant to the requirements of 10 CFR 50.73. 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 revise organizational and administrative
requirements contained within the Administrative Controls section of
the TS. The proposed change to the definition of Azimuthal Power
Tilt is as recommended in CE Infobulletin 97-07 for CE analog plants
that use CECOR for monitoring and surveillance purposes and will
have no affect on accidents previously evaluated. The proposed
changes do not revise any equipment setpoints, change the manner in
which any plant equipment is operated, or propose any new operating
modes. Therefore, the proposed changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes revise organizational and administrative
requirements contained within the Administrative Controls section of
the TS. The proposed change to the definition of Azimuthal Power
Tilt has no affect on the margin of safety. The proposed changes do
not revise any equipment setpoints, change the manner in which any
plant equipment is
[[Page 34288]]
operated, or propose any new operating modes. 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: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: May 17, 2001.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to permit an increase in the
allowable leak rate for the main steam isolation valves (MSIVs) and to
delete the MSIV Sealing System (MSIVSS). These changes are based on the
use of an alternate source term and the guidance provided in Regulatory
Guide 1.183, ``Alternate Radiological Source Terms for Evaluating
Design Basis Accidents at Nuclear Power Reactors.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff's review is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
As described in Section 6.7 of the Hope Creek Updated Final
Safety Analysis Report (UFSAR), the MSIVSS limits the leakage of
fission products through the MSIVs following a design-basis accident
large break Loss of Coolant Accident (LOCA). The system is manually
actuated following a LOCA. The licensee has proposed to remove the
MSIVSS from the plant and to delete the associated requirements from
the TSs. In addition, the TSs would be revised to increase the
allowable MSIV leak rate. The MSIVSS lines and main steamline drain
valves that are connected to the main steam piping will be capped
and welded closed to ensure primary containment integrity is
maintained. The welding and post-weld examination procedures will be
in accordance with the American Society of Mechanical Engineers
Code, Section III requirements. The welded caps will be periodically
tested as part of the Containment Integrated Leak Rate Test. MSIV
leakage and operation of the MSIVSS do not affect the precursors for
accidents analyzed in Chapter 15 of the Hope Creek UFSAR. In
addition, the proposed changes do not adversely affect other
structures, systems, or components important to safety. Therefore,
there is no increase in the probability of occurrence of an accident
previously evaluated as a result of the proposed changes.
The licensee's submittal states that the radiological
consequences associated with the proposed changes have been analyzed
based on the results of revised offsite and control room operator
dose calculations for a LOCA, which is the most limiting Hope Creek
design-basis accident. The current design-basis analysis for the
radiological consequences associated with a LOCA is shown in Hope
Creek UFSAR Sections 6.4.7 and 15.6.5.5. The revised analysis was
performed using an alternate source term in accordance with the
requirements in 10 CFR 50.67 and the guidance in Regulatory Guide
1.183. The dose calculations assess the effects of the proposed
increase in allowable MSIV leak rate and take no credit for the
MSIVSS. In addition, the calculations assume an unfiltered control
room inleakage design-basis value that is higher than the current
design basis value to address control room habitability issues
associated with NEI 99-03. The revised analysis was performed in
accordance with the current accepted methodology discussed in
Regulatory Guide 1.183 and the radiological consequences were
evaluated in terms of Total Effective Dose Equivalent (TEDE) dose as
per the acceptance criteria specified in 10 CFR 50.67. The
Regulatory Guide 1.183 methodology is not exactly comparable to the
current Hope Creek design basis analysis which is in terms of whole
body and thyroid doses. The results of the licensee's analysis
associated with the proposed changes indicate that the post-LOCA
doses will result in an increase in the dose exposures for the
control room, the Exclusion Area Boundary (EAB), and the Low
Population Zone (LPZ), compared to the current design basis
analysis. However, the revised post-LOCA doses will remain below the
TEDE dose acceptance criteria for the control room, EAB, and LPZ, as
specified in 10 CFR 50.67. The methodology and guidance provided in
Regulatory Guide 1.183 has been developed for the purpose of
performing design basis radiological consequence analyses using an
alternate source term such that meeting the 10 CFR 50.67 acceptance
criteria demonstrates adequate protection of public health and
safety. Therefore, the proposed changes do not involve a significant
increase in the consequences of any accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change to increase the allowed MSIV leakage rate
does not affect the operability the MSIVs and will not inhibit the
capability of the MSIVs to perform their function of isolating the
primary containment as assumed in the Hope Creek accident analyses
in UFSAR Chapter 15. The proposed change to delete the MSIVSS does
not introduce any new modes of plant operation and, as previously
discussed, the design-basis LOCA analysis was reanalyzed without
taking credit for the operation of MSIVSS. The affected main steam
piping will be welded and/or capped closed to assure that the
primary containment integrity, isolation, and leak testing
capability are not compromised. Based on the above considerations,
the proposed changes do not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
As previously discussed, the results of the licensee's analysis
associated with the proposed changes indicate that the post-LOCA
doses will result in an increase in the dose exposures for the
control room, the EAB, and the LPZ, compared to the current design
basis analysis. Since there will be an increase in dose exposure,
the margin of safety will be decreased. However, the revised post-
LOCA doses will remain below the TEDE dose acceptance criteria for
the control room, EAB, and LPZ, as specified in 10 CFR 50.67.
Meeting the 10 CFR 50.67 acceptance criteria demonstrates adequate
protection of public health and safety. An acceptable margin of
safety is inherent in these acceptance criteria. Therefore, there is
no significant reduction in the margin of safety as a result of the
proposed changes.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: March 5, 2001.
Description of amendment request: The proposed amendment would (1)
change the Security Plan provision that a member of the security force
escort all vehicles, other than designated licensee vehicles, and to
delete the related Security Training and Qualification Plan task, (2)
change the requirement of the Security Plan that all areas of the
protected area be illuminated to a minimum of 0.2 footcandle, and (3)
change the frequency of protected area patrols in the Security Plan.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or
[[Page 34289]]
consequences of an accident previously evaluated?
The proposed changes involving security activities do not reduce
the ability for the security organization to prevent radiological
sabotage and therefore do not increase the probability or
consequences of a radiological release previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed changes involve functions of the security
organization concerning vehicle control, protected area
illumination, and protected area patrol frequency. Analysis of the
proposed changes has not indicated nor identified 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?
Analysis of the proposed changes show that they affect only the
functions of the Security organization and have no impact upon nor
cause a significant reduction in margin of safety for plant
operation. The failure points of key safety parameters are not
affected.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: May 30, 2001 (ULNRC-04481).
Description of amendment request: The proposed amendment changes
the technical specifications to remove the phrase ``and the charging
flow control valve full open'' from Limiting Condition for Operation
3.5.5, Required Action A.1, and Surveillance Requirement 3.5.5.1 for
the reactor coolant pump seal injection flow.
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 emergency core cooling system (ECCS) analysis models the
reactor coolant pump (RCP) seal injection flow path as a hydraulic
flow resistance. The proposed change clarifies that RCP seal
injection flow is a function of system conditions. The seal
injection flow rate can vary during operation, but the hydraulic
flow resistance is fixed by positioning the manual seal injection
throttle valves. The resistance does not change if the valve
adjustments are not changed. Thus, RCP seal injection flow variation
due to changing reactor coolant system (RCS) backpressure following
a loss of coolant accident (LOCA) is explicitly accounted for as a
result of modeling the RCP seal injection flow path resistance.
The proposed change does not impact the way the RCP seal
injection flow should be established per the safety analysis and
does not affect RCP seal integrity. The seal injection flow
resistance only affects ECCS flow. Since ECCS flow occurs after an
accident, the proposed change cannot impact the probability of an
accident.
Overall ECCS performance will remain within the bounds of the
previously performed accident analyses since there are no hardware
changes. The ECCS will continue to function in a manner consistent
with the plant design basis. All design, material, and construction
standards that were applicable prior to the proposed change are
[still] maintained.
The proposed change will not affect the probability of any event
initiators. There will be no degradation in the performance of, or
an increase in the number of challenges imposed on, safety-related
equipment assumed to function during an accident situation. There
will be no change to normal plant operating parameters or accident
mitigation performance.
The proposed change will not alter any assumptions or change any
mitigation actions in the radiological consequence evaluations in
the FSAR [Final Safety Analysis Report].
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There are no hardware changes nor are there any changes in the
method by which any safety-related plant system performs its safety
function. The proposed change will not affect the normal method of
plant operation. No performance requirements will be affected.
Since the proposed change continues to assure that the assumed
ECCS flow is available after a large break LOCA, no new accident
scenarios, transient precursors, failure mechanisms, or limiting
single failures are introduced as a result [of the proposed change].
There will be no adverse effect or challenges imposed on any safety-
related system as a result of this request.
The proposed change does not alter the design or performance
characteristics of the ECCS. It simply corrects the description of
how to properly set the position of the RCP seal injection throttle
valves in support of the ECCS flow balance assumptions.
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.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions. There will be no impact on the overpower
limit, departure from nucleate boiling ratio limits, heat flux hot
channel factor (FQ) nuclear enthalpy rise hot channel
factor (FN/DH), loss of coolant accident peak cladding temperature
(LOCA PCT), peak local power density, or any other margin of safety.
The radiological dose consequence acceptance criteria listed in the
Standard Review Plan will continue to be met.
Therefore, the proposed change does not involve a significant
reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: April 11, 2000, as supplemented by
letters dated August 28, 2000, November 20, 2000, and April 11, 2001.
Description of amendment request: The proposed amendments would
revise Technical Specifications (TS) 3.7, 3.10, and 3.22, as well as
the Bases of TS 3.4, 3.8, 3.10, 3.19, and 3.22. The proposed changes
would implement an alternate accident source term methodology
previously approved by NRC. Implementation of the alternate source term
could permit a number of plant changes that have been proposed,
including: Permitting a slight atmospheric pressure in containment for
a short time following a loss-of-coolant accident (LOCA), deletion of
automatic function requirements and setpoints for containment
particulate and gas monitors, deletion of the requirement to filter
fuel building and containment purge exhaust during refueling, and a
number of other related operational and configuration requirements.
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:
[[Page 34290]]
The proposed TS changes allow relaxation of containment
integrity requirements during refueling operations by allowing the
personnel airlock, equipment access hatch and certain penetrations
to remain open during fuel movement in containment. The changes also
eliminate the requirement to filter the exhaust from containment or
the fuel building during refueling operations. Also proposed is a
relaxation of the current containment design basis acceptance
criteria to allow an interval of four hours following the design
basis LOCA until containment is depressurized to subatmospheric
conditions. We have reviewed the proposed TS changes relative to the
requirements of 10 CFR 50.92 and determined that a significant
hazards consideration is not involved. Specifically, operation of
Surry Power Station with the proposed changes will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The probability remains unaffected since the accident analyses
involve no change to a system, component or structure that affects
initiating events for any of the accidents evaluated. The
consequences of the reanalyzed events is expressed in terms of the
TEDE [total effective dose equivalent] dose, which is not directly
comparable to either the thyroid or whole body doses reported in
existing analyses. However, even taking this comparison into
consideration, any dose increase is not significant. Furthermore,
the revised analysis results meet the applicable TEDE dose
acceptance criteria for alternative source term implementation.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The implementation of the proposed changes does not create the
possibility of an accident of a different type than was previously
evaluated in the SAR [Safety Analysis Report]. The proposed
Technical Specifications changes allow relaxation of these current
requirements: (1) maintaining subatmospheric containment conditions
following a LOCA; (2) filtration of containment & fuel building
exhaust during fuel movement; (3) maintaining the personnel airlock,
equipment access hatch & penetrations closed during fuel movement
and (4) operability of containment purge isolation during refueling.
These changes do not alter the nature of events postulated in the
UFSAR [Updated Final SAR] nor do they introduce any unique precursor
mechanisms. Therefore, there is no possibility for accidents of a
different type than previously evaluated.
3. Involve a significant reduction in the margin of safety.
The implementation of the proposed changes does not reduce the
margin of safety. The radiological analysis results, even though
compared with the revised TEDE acceptance criteria, meet the
applicable limits. These criteria have been developed for
application to analyses performed with alternative source terms.
These acceptance criteria have been developed for the purpose of use
in design basis accident analyses such that meeting the stated
limits demonstrates adequate protection of public health and safety.
It is thus concluded that the margin of safety will not be reduced
by the implementation of the changes.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams,
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia
23219.
NRC Section Chief: Richard L. Emch.
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].
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of application for amendments: November 30, 2000.
Brief description of amendments: The amendments revise TS 5.5.13,
``Diesel Fuel Oil Testing Program,'' to relocate the specific American
Society for Testing and Materials (ASTM) Standard reference from the
Administrative Controls Section of TS to a licensee-controlled
document, i.e., the Diesel Fuel Oil Program in the Technical
Requirements Manual (TRM). In addition, the ``clear and bright'' test
used to establish the acceptability of new fuel oil for use prior to
addition to storage tanks has been expanded to allow a water and
sediment content test to be performed to establish the acceptability of
new fuel oil in lieu of the ``clear and bright'' test.
Date of issuance: June 13, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 122, 122, 116, and 116.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 21, 2001.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 13, 2001.
No significant hazards consideration comments received: No.
Exelon Generation Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: July 31, 2000.
Brief description of amendments: Revised Technical Specification
(TS) Surveillance Requirement (SR) 4.5.1.d.1, concerning the
operability of the Automatic Depressurization System, and relocated the
existing requirements
[[Page 34291]]
in TS SR 4.5.1.d.1 and TS SR 4.5.1.d.2.c to the Technical Requirements
Manual.
Date of issuance: June 12, 2001.
Effective date: As of date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 152 and 116.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 18, 2000 (65 FR
62389).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 12, 2001.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334,
Beaver Valley Power Station, Unit No. 1, Beaver County, Pennsylvania
Date of application for amendment: November 8, 2000, as
supplemented on February 6, and May 7, 2001.
Brief description of amendment: The amendment changed the technical
specifications associated with the deletion of TS 3/4.4.1.6, ``Reactor
Coolant Pump--Startup.''
Date of issuance: June 13, 2001.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No: 238.
Facility Operating License No. DPR-66: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 27, 2000 (65
FR 81917).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 13, 2001.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit 2, Berrien County, Michigan
Date of application for amendment: January 19, 2001, as
supplemented April 20 and May 9, 2001.
Brief description of amendment: The amendment would change the TSs
to extend surveillance intervals associated with the emergency diesel
generator (EDG) engines and station batteries that are currently
required to be completed beginning June 27, 2001. The license amendment
would allow these requirements to be performed during the next
refueling outage, but no later than December 31, 2001. This would
preclude the need for a mid-cycle shutdown of the Unit.
Date of issuance: June 11, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 234.
Facility Operating License No. DPR-74: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 21, 2001 (66 FR
15926). The April 20 and May 9, 2001, supplemental letters, did not
change the scope of the proposed action and did not change the Nuclear
Regulatory Commission's (NRC's) preliminary no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 11, 2001.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: January 13, 2000, as
supplemented March 7, March 30, and May 4, 2001.
Brief description of amendment: The amendment revises the Kewaunee
Nuclear Power Plant (KNPP) Technical Specifications (TSs) 3.6,
``Containment'' to add Limiting Condition for Operation (LCO) and
Allowed Outage Times (AOT) for containment isolation devices. In
addition, the amendment provides additional information, clarification,
and uniformity to the bases of the associated TSs.
Date of issuance: June 8, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 155.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 21, 2001 (66
FR 11061). The March 7, March 30, and May 4, 2001, letters, provided
clarifying information that was within the scope of the original
application, did not change the NRC staff's initial proposed no
significant hazards consideration determination, and did not expand the
amendment beyond the scope of the original notice (66 FR 11061).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 8, 2001.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of application for amendment: January 26, 2001, as
supplemented by letter dated March 13, 2001.
Brief description of amendment: The amendment changes Technical
Specification Surveillance Requirement 3.7.9.2, ``Ultimate Heat Sink
(UHS),'' by increasing the maximum allowable temperature of Lake
Michigan water from 81.5 deg.F to 85 deg.F.
Date of issuance: June 4, 2001.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 202.
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 7, 2001 (66 FR
13800).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 4, 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: August 3, 2000, as supplemented by
letters dated November 17, 2000, and February 14, 2001.
Brief description of amendment: The amendment deletes Section 3.D,
``License Term,'' from the Fort Calhoun Station, Unit No. 1 operating
license.
Date of issuance: June 6, 2001.
Effective date: June 6, 2001, to be implemented within 30 days from
the date of issuance.
Amendment No.: 199.
Facility Operating License No. DPR-40: The amendment revised the
operating license.
Date of initial notice in Federal Register: January 10, 2001 (66 FR
2019).
The November 17, 2000, and February 14, 2001, supplemental letters
provided clarifying information, 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 June 6, 2001.
No significant hazards consideration comments received: No.
[[Page 34292]]
Southern California EdisonCompany, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: April 6, 2001 and supplemented
by letter dated April 20, 2001.
Brief description of amendments: The amendments proposed to revise
the San Onofre Nuclear Generating Station, Units 2 and 3 Technical
Specification Surveillance Requirements 3.8.1.2, 3.8.1.3, 3.8.1.9,
3.8.1.10, and 3.8.1.19 to assure that an emergency diesel generator
automatic voltage regulator (AVR) is operable and regularly tested. AVR
operability would be demonstrated by conducting SR 3.8.1.2 and 3.8.1.3
within the past 60 days, and any one of SR 3.8.1.9, 3.8.1.10, or
3.8.1.19 within the past 24 months.
Date of issuance: June 8, 2001.
Effective date: June 8, 2001, to be implemented within 30 days of
issuance.
Amendment Nos.: Unit 2-179; Unit 3-170.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 2, 2001 (66 FR
22032).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 8, 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: August 17, 2000, as supplemented by
letter dated April 2, 2001. The April 2, 2001, letter requested a new
implementation date, but did not change the August 17, 2000,
application and the initial proposed no significant hazards
consideration determination.
Brief description of amendments: The amendments eliminate the need
for the licensee to perform periodic response time testing of selected
reactor trip system and engineered safety feature actuation system
equipment as defined in Westinghouse report WCAP-14036-P-A, Revision 1,
``Elimination of Periodic Protection Channel Response Time Tests.''
Date of issuance: June 7, 2001.
Effective date: As of the date of issuance and shall be implemented
on Unit 1 entry in Mode 3 for Cycle 18 following the 2001 fall
refueling.
Amendment Nos.: 149 and 141.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: January 10, 2001 (66 FR
2023). The supplement dated April 2, 2001, provided clarifying
information that did not change the scope of the August 17, 2001,
application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 7, 2001.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of application for amendments: January 11, 2001.
Brief description of amendments: The amendments revise TS 5.5.17,
``Containment Leakage Rate Testing Program,'' to add an exception to
Regulatory Guide 1.163 related to visual examination of containment
concrete surfaces.
Date of issuance: June 6, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 122 and 100.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 2, 2001 (66 FR
22033).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 6, 2001.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant , Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: November 6, 2000.
Description of amendment request: These amendments revised the
Technical Specifications (TS) to allow four residual heat removal
suppression pool cooling subsystems to be inoperable for 8 hours.
Date of issuance: June 8, 2001.
Effective date: June 8, 2001.
Amendment Nos.: 241, 272, and 230.
Facility Operating License Nos. DPR-33, DPR-52, and DPR-68.
Amendments revised the TS.
Date of initial notice in Federal Register: November 29, 2000 (65
FR 71139).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 8, 2001.
No significant hazards consideration comments received: No.
TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: April 3, 2001.
Brief description of amendments: The amendments revise Technical
Specification (TS) 3.3.6, ``Containment Ventilation Isolation
Instrumentation,'' to modify the Note for Required Action B.1 such that
it applies only to * * * Required Action and associated Completion Time
of Condition A not met * * * This change is the result of the discovery
of an error which occurred when the TSs were converted to the improved
TS with issuance of License Amendment Nos. 64 and 64, for Comanche Peak
Steam Electric Station, Units 1 and 2, on February 26, 1999.
Date of issuance: June 4, 2001.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 86 and 86.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: May 2, 2001 (66 FR
22034).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 4, 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: September 27, 2000, as
supplemented November 21 and December 18, 2000, and February 2, March
2, and May 21, 2001.
Brief description of amendment: These amendments add Technical
Specification (TS) 3.7.14, TS 4.7.14, TS 3.7.15, TS 4.7.15, Figure
3.7.15-1, and Figure 3.7.15-2; and revise TS 5.3.1 and TS 5.6.1.1. The
purpose of these amendments is to increase the limit on the fuel
enrichment from the current limit of 4.3 weight percent U235
to a maximum of 4.6 weight percent U235, establish TS
Limiting Conditions for Operations for the Spent Fuel Pool (SFP) boron
concentration and fuel storage restrictions, and eliminate the value of
uncertainties in the calculation for Keff in the SFP
criticality calculation.
[[Page 34293]]
Date of issuance: June 15, 2001.
Effective date: As of the date of issuance and shall be implemented
by December 21, 2001.
Amendment Nos.: 227 and 208.
Facility Operating License Nos. NPF-4 and NPF-7: Amendments change
the Technical Specifications.
Date of initial notice in Federal Register: December 13, 2000 (65
FR 77929). The December 18, 2000, February 2, March 2, and May 21,
2001, supplements contained clarifying information only, and did not
change the initial no significant hazards consideration determination,
or expand the scope of the initial application.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 15, 2001.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 19th day of June 2001.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 01-15818 Filed 6-26-01; 8:45 am]
BILLING CODE 7590-01-P