[Federal Register Volume 66, Number 192 (Wednesday, October 3, 2001)]
[Notices]
[Pages 50463-50480]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-24580]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
Note: The publication date for this notice will change from
every other Wednesday to every other Tuesday, effective January 8,
2002. The notice will contain the same information and will continue
to be published biweekly.
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from September 10, 2001 through September 21,
2001. The last biweekly notice was published on September 19, 2001 (66
FR 48283).
Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC Public Document
Room, located at One White Flint North, 11555 Rockville Pike (first
floor), Rockville, Maryland. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By November 2, 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
[[Page 50464]]
Domestic Licensing Proceedings'' in 10 CFR part 2. Interested persons
should consult a current copy of 10 CFR 2.714, which is available at
the NRC's Public Document Room, located at One White Flint North, 11555
Rockville Pike (first floor), Rockville, Maryland 20852. Publicly
available records will be accessible electronically from the Agencywide
Documents Access and Management Systems (ADAMS) Public Electronic
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/
NRC/ADAMS/index.html. If a request for a hearing or petition for leave
to intervene is filed by the above date, the Commission or an Atomic
Safety and Licensing Board, designated by the Commission or by the
Chairman of the Atomic Safety and Licensing Board Panel, will rule on
the request and/or petition; and the Secretary or the designated Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Branch, or may be delivered to the Commission's Public
Document Room, located at One White Flint North, 11555 Rockville Pike
(first floor), Rockville, Maryland 20852, by the above date. A copy of
the petition should also be sent to the Office of the General Counsel,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and to
the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Assess and Management Systems (ADAMS) Public Electronic
Reading Room on the internet at the NRC Web site, http://www.nrc.gov/
NRC/ADAMS/index.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC Public Document room (PDR) Reference staff at 1-800-397-4209, 304-
415-4737 or by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: June 21, 2001 (U-603490)
Description of amendment request: The proposed amendment would
eliminate the Technical Specification leakage limit for any one main
steam line as measured by main steam isolation valve leakage of less
than or equal to 28 standard cubic feet per hour (scfh) and replace
that requirement with an aggregate leakage limit of less than or equal
to 112 scfh for all four main steam lines.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The MSIVs [main steam isolation valves] are not initiators of or
precursors to any of the accident scenarios presented in the Updated
Safety Analysis Report. Therefore, this change does not involve an
increase in the probability of any accident previously evaluated.
The proposed change to the TS [Technical Specifications]
modifies the allowed main steam line leakage limit to an aggregate
value (i.e., leakage for all four main steam lines combined) with no
change to the currently allowed total leakage rate. This is the
value currently used for calculation of dose consequences for the
bounding accident for which MSIV closure is credited, the large-
break loss of coolant accident (LOCA). This proposed change does not
impact or increase
[[Page 50465]]
the assumed radionuclide source term therefore; this change does not
involve an increase in consequences of any accident previously
evaluated.
In summary, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve a change to the plant
design or operation. No new equipment will be installed or utilized,
and no new operating conditions will be initiated as a result of
this change. The safety function of the MSIVs is to provide timely
steam line isolation to mitigate the release of radioactive steam
and limit reactor inventory loss under certain accident and
transient conditions. Changing the leakage limits to include an
aggregate value does not affect the isolation function performed by
the MSIVs. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
The total allowed leakage rate for all four main steam lines
remains unchanged at 112 scfh. The proposed change does
not challenge the integrity of the fuel cladding, reactor coolant
pressure boundary, or the primary containment.
The margin of safety that has the potential of being impacted by
the proposed change involves the offsite dose consequences of
postulated accidents which are directly related to containment
leakage. As stated above, the total allowed leakage rate for all
four main steam lines remains unchanged. In addition, there will not
be a change in the types or amounts of any effluents released
offsite. The radiological analyses remain unchanged and within the
guidelines of 10 CFR 100, ``Reactor Site Criteria,'' and 10 CFR 50,
``Domestic Licensing of Production and Utilization Facilities,''
Appendix A, ``General Design Criteria for Nuclear Power Plants,''
General Design Criterion 19, ``Control Room.''
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: Robert Helfrich, Mid-West Regional Operating
Group, Exelon Generation Company, LLC, 1400 Opus Place, Suite 900,
Downers Grove, IL 60515.
NRC Section Chief: Anthony J. Mendiola.
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: June 21, 2001 (U-603495).
Description of amendment request: The proposed amendment would
modify the Technical Specification requirement that the main steam line
safety relief valves (SRVs) open when they are manually actuated by
instead requiring that the SRV valve actuators stroke on a manual
actuation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes modify TS [Technical Specification] SR
[surveillance requirements] 3.4.4.3, SR 3.5.1.7 and SR 3.6.1.6.1.
The proposed changes will eliminate the TS requirement that each
valve opens during the manual actuation of the SRVs [safety relief
valves]. Accidents are initiated by the malfunction of plant
equipment, or the catastrophic failure of plant structures, systems
or components. The performance of SRV testing is not a precursor to
any accident previously evaluated and does not change the manner in
which the SRVs are operated. The proposed testing requirements will
not contribute to the failure of the SRVs nor any plant structure,
system or component. Thus, the proposed changes to the performance
of SR 3.4.4.3, SR 3.5.1.7 and SR 3.6.1.6.1 do not have any affect on
the probability of an accident previously evaluated.
The performance of SRV testing provides assurance that the SRVs
are capable of depressurizing the reactor pressure vessel (RPV).
This will protect the reactor vessel from overpressurization and
allowing the combination of the Low Pressure Coolant Injection
(LPCI) System and Low Pressure Core Spray (LPCS) System to inject
into the RPV as designed. The LLS [low-low set] logic causes two LLS
valves to be opened at a lower pressure than the relief or safety
mode pressure setpoints and causes all the LLS valves to stay open
longer, such that reopening of more than one SRV is prevented on
subsequent actuations. Thus, the LLS function prevents excessive
short duration SRV cycles with valve actuation at the relief
setpoint. The proposed changes involve the manner in which the
subject valves are tested, and have no affect on the types or
amounts of radiation released or the predicted offsite doses in the
event of an accident. The proposed testing requirements are
sufficient to provide confidence that the SRVs, ADS valves and the
LLS valves will perform their intended safety functions. Thus, the
radiological consequences of any accident previously evaluated are
not increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes to SR 3.4.4.3, SR 3.5.1.7 and SR 3.6.1.6.1
do not affect the assumed accident performance of the SRVs, nor any
plant structure, system or component previously evaluated. The
proposed changes do not install any new equipment, and installed
equipment is not being operated in a new or different manner. The
valves continue to be bench-tested to verify the safety and relief
modes of valve operation. The proposed changes will allow the
testing of the manual actuation electrical circuitry, solenoid and
air control valve, and the actuator without causing the SRV to open.
No setpoints are being changed which would alter the dynamic
response of plant equipment. Administrative controls, such as
verifying that the actuator assembly has been recoupled following
testing, minimize the potential for valve failures. Accordingly, no
new failure modes are introduced. The changes credit the performance
of bench testing, setpoint verification and in-situ actuator
exercising with providing sufficient testing to ensure the valves
will perform their required safety functions.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
Does the change involve a significant reduction in a margin of
safety?
The proposed changes to SR 3.4.4.3, SR 3.5.1.7 and SR 3.6.1.6.1
will allow the uncoupling of the SRV stem from the other components
associated with the manual actuation of the SRVs. The proposed
changes will allow the testing of the manual actuation electrical
circuitry, solenoid and air control valve, and the actuator without
causing the SRV to open. The SRVs will continue to be manually
actuated by the bench-test valve control system of the setpoint
testing program and prior to installation in the plant. The proposed
changes do not effect the valve setpoint or the operational criteria
that directs the SRVs to be manually opened during plant transients.
There are no changes proposed which alter the setpoints at which
protective actions are initiated, and there is no change to the
operability requirements for equipment assumed to operate for
accident mitigation.
Thus, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c)
are satisfied. Therefore, the NRC staff proposes to determine that
the amendment request involves no significant hazards consideration.
Attorney for licensee: Robert Helfrich, Mid-West Regional
Operating Group, Exelon Generation Company, LLC, 1400 Opus Place,
Suite 900, Downers Grove, IL 60515.
NRC Section Chief: Anthony J. Mendiola.
[[Page 50466]]
Duke Energy Corporation, Docket No. 50-287, Oconee Nuclear Station,
Unit 3, Oconee County, South Carolina
Date of amendment request: March 5, 2001, supplemented September
4, 2001
Description of amendment request: The proposed amendments would
revise the Technical Specifications for Unit 3 to allow a one-time
extension of the 10 CFR Part 50, Appendix J Containment Integrated
Leak Rate Test interval. Presently the 10-year interval test is
required to be performed prior to the operating cycle before the
outage when the steam generators will be replaced. The proposed
amendment would extend the test approximately 16 months to the
outage when they will be replaced (i.e., no later than April 11,
2005), thereby precluding the need to perform the test during two
subsequent outages. The No Significant Hazards Consideration
Determination contained in the March 5, 2001, submittal was
superceded in the September 4, 2001, submittal and is presented
below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No
The proposed revision to the Oconee Nuclear Station, Unit 3 (ONS-3)
Technical Specifications (TS) adds a one-time extension to the current
interval for Type A testing (containment Integrated Leak Rate Testing
(ILRT)). The current test interval of 10 years, would be extended on a
one time basis to 12 years 7 months from the last Type A test. The
proposed extension to Type A testing cannot increase the probability of
an accident previously evaluated since the containment Type A testing
extension is not a modification to plant systems, or a change to plant
operation that could initiate an accident. The proposed extension to
Type A testing does not involve a significant increase in the
consequences of an accident since research documented in NUREG-1493
found that, generically, very few potential containment leakage paths
fail to be identified by Type B and C tests. In fact, an analysis of
144 ILRT results, including 23 failures, found that no failures were
due to containment liner breach. The NUREG concluded that reducing the
Type A testing frequency to one per twenty years would lead to an
imperceptible increase in risk. The NUREG conclusions are supported by
an ONS-3 specific evaluation of risk and consequences. ONS-3 provides a
high degree of assurance through testing and inspection that the
containment will not degrade in a manner detectable only by Type A
testing. Inspections required by the Maintenance Rule and American
Society of Mechanical Engineers (ASME) code are performed in order to
identify indications of containment degradation that could affect leak
tightness. Type B and C testing required by the ONS-3 TS will identify
any containment opening, such as valves, that would otherwise be
detected by the Type A tests. Type B and C testing is performed at the
frequency specified by 10 CFR 50, Appendix J, Option A, Sec. D.2 and
Sec. D.3, respectively. These factors show that a ONS-3 Type A test
extension will not represent a significant increase in the consequences
of an accident.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No
The proposed extension to Type A testing cannot create the
possibility of a new or different type of accident since there are no
physical changes [b]eing made to the plant. There are no changes to the
operation of the plant that could introduce a new failure mode creating
the possibility of a new or different kind of accident.
3. Do the proposed changes involve a significant reduction in the
margin of safety?
Response: No
The proposed extension to Type A testing will not significantly
reduce the margin of safety. The NUREG-1493 generic study of the
effects of extending containment leakage testing found that a 20 year
extension in Type A leakage testing resulted in an imperceptible
increase in risk to the public. NUREG-1493 found that, generically, the
design containment leakage rate contributes a very small amount to the
individual risk, and that the decrease in Type A testing frequency
would have a minimal affect on this risk since most potential leakage
paths are detected by Type C testing. The NUREG conclusions are
supported by an ONS-3 specific evaluation of risk and consequences.
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: Anne W. Cottington, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20005.
NRC Section Chief: Richard L. Emch, Jr.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: April 24, 2001.
Description of amendment request: The proposed amendment would
revise the Indian Point 3 (IP3) Final Safety Analysis Report (FSAR)
to reflect the original plant design. It will indicate that a
portion of one loop of the Component Cooling Water (CCW) System is
routed in the non-safety-related portion of the Fuel Storage
Building (FSB).
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:
Operation of the Indian Point 3 plant in accordance with the
proposed amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92 since it would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to the FSAR revises it to reflect the as built
configuration of the Component Cooling Water (CCW) system. One loop is
routed to the Spent Fuel Pit Heat Exchanger (SFPHX) located in the Fuel
Storage Building (FSB). The portion of the FSB where the CCW is located
is seismic Class III rather than the seismic Class I required by design
criteria in the FSAR. The proposed change demonstrates that the CCW
loop and SFPHX will not be affected by a seismic event and that
operator, action with credit for the Primary Water Storage Tank (PWST)
providing redundancy (a source of water to maintain CCW), will assure
that the CCW system function can be performed following a tornado. The
proposed change does not affect the probability of an accident
previously evaluated because there is no design change and the
probability of natural phenomena does not change. The proposed change
does not affect the consequences of an accident previously evaluated
because the CCW system function is maintained by operator action
following a tornado with missile damage to a small bore CCW pipe.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change to the FSAR revises it to reflect the as built
configuration of the CCW system. One loop is routed to the Spent Fuel
Pit Heat Exchanger (SFPHX) located in the Fuel Storage Building (FSB).
The portion of the FSB where the CCW is located is seismic Class III
rather than the seismic Class I required by design criteria in the
FSAR. The proposed change demonstrates that the CCW loop and SFPHX will
not be affected by a seismic event and that operator action with credit
for the PWST inventory (a source of water to maintain CCW) will assure
[[Page 50467]]
that the CCW system function can be performed following a tornado. The
proposed change does not create the possibility of a new or different
kind of accident because the CCW system will not be operated
differently than designed and operator action and the use of components
to perform redundant functions to cope with a tornado is currently
approved in the FSAR. Also, the CCW is designed to be operated by
separating the loops.
3. Involve a significant reduction in a margin of safety.
The proposed change to the FSAR revises it to reflect the as built
configuration of the CCW system. One loop is routed to the Spent Fuel
Pit Heat Exchanger (SFPHX) located in the Fuel Storage Building (FSB).
The portion of the FSB where the CCW is located is seismic Class III
rather than the seismic Class I required by design criteria in the
FSAR. The proposed change demonstrates that the CCW loop and SFPHX will
not be affected by a seismic event and that operator action with credit
for the PWST inventory (a source of water to maintain CCW) will assure
that the CCW system function can be performed following a tornado. The
proposed change does not involve a significant reduction in a margin of
safety because the existing plant design considers the use of operator
action and redundant components to mitigate the effects of a tornado.
Also, the proposed change is for damage caused by a low risk event. The
risk of tornado damage to the CCW piping in the FSB is low. The IP3
examination of external events found the probability of any tornado
striking IP3 to be 1.59E-4/year. For tornados with wind speeds in
excess of 180 mph, the frequency decreases to 8.62E-7/year. For the
design basis tornado with a 300 mph wind speed, the frequency is 1.02E-
9/year. The risk of a tornado following a LOCA is lower. The frequency
of a LOCA followed by any tornado within 30 days is 3.02E-8/year. The
frequency of the event can be used as a conservative estimate of core
damage frequency (CDF). When compared to the nominal CDF at IP3, the
frequency is negligible.
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 Fulton, Assistant General Counsel,
Entergy Nuclear Generating Station, 600 Rocky Hill Road, Plymouth, MA
02360.
NRC Section Chief: Peter Tam, Acting.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois; Docket
Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle
County, Illinois; Docket Nos. 50-237 and 50-249, Dresden Nuclear Power
Station, Units 2 and 3, Grundy County, Illinois; Docket Nos. 50-373 and
50-374, LaSalle County Station, Units 1 and 2, LaSalle County,
Illinois; Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power
Station, Units 1 and 2, Rock Island County, Illinois
Date of amendment request: March 23, 2001.
Description of amendment request: The proposed amendment would
incorporate changes to the Physical Security Plan and Guard Force
Training and Qualification Plans for the identified facilities. The
proposed changes would modify current escorting and control
requirements for non-designated vehicles, lighting requirements for
exterior areas within the protected area, and annual weapons
qualifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
No physical plant changes are being made as a result of changing
the vehicle, lighting, and weapons qualification requirements. The
proposed changes involve revising requirements that provide little
or no value in the protection of the facility with regards to the
design basis threat as described in 10 CFR part 73, ``Physical
Protection of Plants and Materials,'' paragraph 1(a). Because the
defensive strategies at each station have been proven to be
effective without reliance on these requirements, it is concluded
that the proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
B. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There are no physical changes being made to the plant as a
result of changing the vehicle, lighting, and weapons qualification
requirements. The defensive strategies at each station remain
unchanged under the proposed changes. A review of possible intrusion
scenarios has confirmed that no event would result in a new sequence
of events that could lead to a new accident scenario. Based on this
review, it is concluded that no accident scenarios, failure
mechanisms or limiting single failures are introduced as a result of
the proposed changes. Therefore, the proposed Physical Security Plan
and Guard Force Training and Qualification Plan changes do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
C. The proposed changes do not involve a significant reduction
in the margin of safety.
It has been shown during recent Operational Safeguards Readiness
Evaluations (OSRE), that the proposed changes do not impact the
security's ability to protect the facility from the threat of
radiological sabotage. The risk of radiological sabotage would not
be increased by changing the vehicle, lighting, and weapons
qualification requirements. Additionally, proposed change in weapons
qualifications provides a more realistic evaluation of a responder's
ability to protect the station from the threat of radiological
sabotage. Based on this review, the proposed amendment does not
involve a significant reduction in the margin of safety.
Therefore, based upon the above evaluation, Exelon Generation
Company, LLC has concluded that these changes do not involve
significant hazards considerations.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Jr., Vice President
and General Counsel, Exelon Generation Company, LLC, 300 Exelon Way,
KSB 3-W, Kennett Square, PA 19348.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), Beaver
County, Pennsylvania
Date of amendment requests: May 22, 2001.
Description of amendment requests: The proposed amendments would
revise the BVPS-1 and 2 technical specifications (TSs) to implement
improvements endorsed in the Nuclear Regulatory Commission's (NRC's)
Final Policy Statement on Technical Specification Improvements for
Nuclear Power Reactors, dated July 22, 1993 (58 FR 39132). These
license amendment requests propose the addition of an administrative
control program for explosive gas and storage tank radioactivity
monitoring to the administrative controls section of the BVPS-1 and 2
TSs consistent with the corresponding standard TS program. The
amendment requests propose to
[[Page 50468]]
relocate the TS requirements associated with the curie content limit
for liquid and gaseous waste storage system and the explosive gas
concentration limit for gaseous waste storage systems. The addition of
the standard TS program provides an appropriate level of control for
the affected requirements in the TSs that allows these details to be
relocated. The TSs proposed for relocation will be placed in the BVPS-1
and 2 Offsite Dose Calculation Manual or the BVPS-1 and 2 Licensing
Requirements Manual. The net effect of the proposed changes is to
provide adequate regulatory control in the TSs while making the content
of the BVPS-1 and 2 TSs more consistent with the standard TSs for
Westinghouse plants as presented in NUREG-1431 and simplifying the
BVPS-1 and 2 TSs consistent with the goals of the NRC Final Policy
Statement on TS improvements for nuclear power reactors.
Additionally, revisions to BVPS-1 and 2 TS 6.9.3, ``Annual
Radioactive Release Report,'' are proposed to include changes to the
reporting requirements. The changes proposed also include various
administrative revisions to support the relocations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed amendment does not involve a significant increase
in the probability of an accident previously evaluated because no
changes are being made to any event initiator. Nor is any analyzed
accident scenario being revised. The initiating conditions and
assumptions for accidents described in the UFSAR [Updated Final
Safety Analysis Report] remain as previously analyzed.
The proposed amendment also does not involve a significant
increase in the consequences of an accident previously evaluated.
The amendment does not reduce the current operability requirements
contained in the TS proposed for relocation. The proposed relocation
of TS requirements only affects the level of regulatory control
involved in future changes to the requirements. The proposed changes
include additions to the TS in the form of programmatic controls
that effectively replace the key TS requirements being relocated. As
such, the TS proposed for relocation no longer meet the 10 CFR 50.36
criteria for retention in the TS.
The additional administrative changes are editorial in nature,
and are made to support the relocation of TS. The additional
administrative changes and the changes to Specification 6.9.3 have
no adverse effect on the safety analyses for design basis accidents
described in the UFSAR. The initiating conditions and assumptions
for accidents described in the UFSAR remain as previously analyzed.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed amendment does not involve any physical changes to
the plant or the modes of plant operation defined in the TS. The
proposed amendment does not involve the addition or modification of
plant equipment nor does it alter the design or operation of any
plant systems. No new accident scenarios, transient precursors,
failure mechanisms, or limiting single failures are introduced as a
result of these changes.
There are no changes in this amendment that would cause the
malfunction of safety-related equipment assumed to be operable in
accident analyses. No new mode of failure has been created and no
new equipment performance requirements are imposed. The proposed
amendment has no effect on any previously evaluated accident.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The margin of safety depends on the maintenance of specific
operating parameters and systems within design requirements and
safety analysis assumptions.
The proposed amendment does not involve revisions to any safety
limits or safety system setting that would adversely impact plant
safety. The proposed amendment does not alter the functional
capabilities assumed in a safety analysis for any system, structure,
or component important to the mitigation and control of design bases
accident conditions within the facility. Nor does this amendment
revise any parameters or operating restrictions that are assumptions
of a design basis accident. In addition, the proposed amendment does
not affect the ability of safety systems to ensure that the facility
can be placed and maintained in a shutdown condition for extended
periods of time.
The proposed change includes the addition of programmatic
controls that allow the affected TS to be relocated. The relocation
of TS does not reduce the effectiveness of the requirements being
relocated. Rather, the relocation of the TS results in a change in
the regulatory control required for future changes made to the
requirements. Additionally, due to the new programmatic controls,
the TS proposed for relocation no longer meet the 10 CFR 50.36
criteria for retention in the TS.
The requirements contained within the affected TS will continue
to be implemented by the appropriate plant procedures (e.g.,
operating and maintenance procedures) in the same manner as before.
However, future changes to the relocated requirements will be
controlled in accordance with 10 CFR 50.59 instead of a license
amendment pursuant to 10 CFR 50.90. The provisions of 10 CFR 50.59
establish adequate controls over requirements removed from the TS
and assure future changes to these requirements will be consistent
with safe plant operation.
The additional administrative changes are editorial in nature,
and are made to support the relocation of TS. The additional
administrative changes and the proposed changes to Specification
6.9.3 do not alter any operating parameters or design requirements
assumed in a safety analysis for systems or components important to
the mitigation and control of design bases accident conditions
within the facility. Nor do these changes alter safety limits or
safety system settings required for safe operation of the plant or
the assumptions of any safety analysis.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Peter Tam (Acting).
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412,
Beaver Valley Power Station, Unit 2 (BVPS-2), Beaver County,
Pennsylvania
Date of amendment request: July 25, 2001.
Description of amendment request: The proposed license amendment
for BVPS-2 would increase the limits for boron concentration in the
refueling water storage tank (RWST) and in the reactor coolant system
(RCS) accumulators. The RCS minimum boron concentration limit for Mode
6 would also be revised to make it consistent with the RWST boron
concentration limit. The increase in the boron concentration limits in
the RWST and accumulators is needed to address higher reactor core
reactivity levels associated with core operation at higher plant
capacity factors. TS Bases changes are also proposed to reflect the
changes discussed above.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
[[Page 50469]]
No. The proposed change to increase the boron concentration in
the Beaver Valley Power Station (BVPS) Unit 2 Refueling Water
Storage Tank (RWST), Accumulators and in the Reactor Coolant System
(RCS) during Mode 6 will maintain the safety analyses results in
Chapter 15 of the BVPS Unit 2 Updated Final Safety Analysis Report
(UFSAR) as bounding values for all Loss Of Coolant Accident (LOCA)
and non-LOCA design basis accidents. The proposed changes do not
reduce the RWST or accumulators ability to meet their design bases,
which will not result in a significant increase in the probability
of an accident previously evaluated.
Increased boron concentration limits for the RWST, Accumulators,
and RCS in Mode 6 will not increase the consequences of an accident
previously analyzed as described in the UFSAR. The increased boron
concentration limits reduce the time to switchover from cold leg to
hot leg recirculation, which will prevent boron precipitation in the
reactor vessel following a LOCA. The post-LOCA long term core
cooling minimum boron requirements have been determined to continue
to be adequate to ensure adequate post-LOCA shutdown margin. The
post-LOCA containment sump and containment spray pH remain within
the limits specified in the UFSAR. All other transients either were
not impacted or were made less severe as a result of the increased
boron concentrations.
Therefore, based upon the above, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. The proposed increase in boron concentration does not add
new or different equipment to the facility. The proposed Technical
Specification changes also do not alter the manner in which plant
equipment is being operated. Although the increased boron
concentration requires procedure changes to ensure that cold leg to
hot leg recirculation after a LOCA occurs quicker, there are no
changes to the methods utilized to respond to plant events. The
proposed Technical Specification changes do not alter instrument or
control setpoints that initiate protective or mitigative actions.
These increased boron concentration limits are conservative and do
not alter the RCS or Emergency Core Cooling Systems' ability to
perform their design bases.
Therefore, these proposed changes do not create the possibility
of a new or different kind of accident from any previously evaluated
accident since the RCS will continue to operate in accordance with
their design bases.
3. Does the change involve a significant reduction in a margin
of safety?
No. The LOCA considerations, including Peak Cladding Temperature
calculations, containment sump and spray pH requirements, boron
solubility requirements, cold shutdown boration requirements, post-
LOCA long term core cooling minimum boron requirements, hot leg
recirculation switchover requirements, post-LOCA hydrogen generation
requirements, and radiological requirements have been evaluated and
determined to be acceptable. The acceptance criteria of all non-LOCA
design basis accidents continue to be met.
The proposed amendment does not involve revisions to any safety
limits or safety system setting that would adversely impact plant
safety. The proposed amendment does not adversely affect the ability
of systems, structures or components important to the mitigation and
control of design bases accident conditions within the facility. In
addition, the proposed amendment does not affect the ability of
safety systems to ensure that the facility can be maintained in a
shutdown or refueling condition for extended periods of time.
Based upon the above evaluations, [the proposed changes do not
involve a significant reduction in a margin of safety.]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Peter Tam, Acting.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: August 22, 2001.
Description of amendment request: The proposed amendments revise
Section 6.0, ``Administrative Controls,'' of the Technical
Specifications (TS) to change the title of the corporate executive
responsible for overall nuclear safety from ``President-Nuclear
Division'' to ``Chief Nuclear Officer.'' The proposed changes eliminate
the reference to a specific organizational title and replace it with a
generic organizational position title. This conforms the TS to a recent
organizational change and precludes the need for future amendments in
response to future corporate title changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) 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, changing
the title of the corporate executive responsible for overall plant
nuclear safety in St. Lucie Units 1 and 2 TS, and would not involve
a significant increase in the probability or consequences of an
accident previously evaluated. These amendments 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,
changing the title of the corporate executive responsible for
overall plant nuclear safety in St. Lucie Units 1 and 2 TS, and
would not create the possibility of a new or different kind of
accident from any previously evaluated. 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. Therefore, the proposed changes do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed changes are administrative in nature, changing the
title of the corporate executive responsible for overall plant
nuclear safety in the St. Lucie Units 1 and 2 TS, and would not
reduce any of the margins of safety.The operating limits and
functional capabilities of the affected systems, structures, and
components remain unchanged by the proposed amendments. 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Richard P. Correia.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: August 15, 2001.
[[Page 50470]]
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) to extend the channel
calibration surveillance frequency for the automatic depressurization
system (ADS) timers from 18 months to 24 months. Specifically, SR
3.3.5.1.7 (18-month CHANNEL CALIBRATION) surveillance requirement
listed in Table 3.3.5.1-1, functions 4.b. and 5.b. (ADS Timer), would
be changed to SR 3.3.5.1.8 (24-month CHANNEL CALIBRATION.) This channel
calibration surveillance would continue to be performed in the same
manner but at a reduced frequency. No modifications to test
methodologies or station equipment have been proposed in this request.
This request is made to facilitate a change to the Duane Arnold Energy
Center operating cycle from 18 months to 24 months. This request has
been prepared following the guidance in Generic Letter 91-04, ``Changes
in Technical Specification Surveillance Intervals to Accommodate a 24-
Month Fuel Cycle.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed amendment extends the CHANNEL CALIBRATION
surveillance frequency for the ADS timers from 18 months to 24
months to facilitate a change in the DAEC operating cycle from 18
months to 24 months. The proposed change does not physically impact
the plant nor does it impact any design or functional requirements
of the ADS. That is, the proposed change does not degrade the
performance or increase the challenges of any safety systems assumed
to function in the accident analysis. The proposed change alters the
frequency but not the Surveillance Requirement itself nor the way in
which the surveillance is performed. The proposed change does not
affect the availability of equipment or systems required to mitigate
the consequences of an accident because of the availability of
redundant systems or equipment and because other tests performed
more frequently will identify potential equipment problems.
Furthermore, an evaluation of surveillance test results shows that
the probability of exceeding the TS Allowable Value (AV) with the
extended surveillance frequency is small and remains well within the
setpoint methodology guideline. Therefore, the proposed amendment
will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed amendment extends the CHANNEL CALIBRATION
surveillance frequency for the ADS timers from 18 months to 24
months to facilitate a change in the DAEC operating cycle from 18
months to 24 months. The proposed change does not introduce any
failure mechanisms of a different type than those previously
evaluated since there are no physical changes being made to the
facility. In addition, only the frequency will change; the
Surveillance Requirement itself and the way the surveillance is
performed will remain unchanged. Furthermore, a review of the
maintenance history of these timers indicated no evidence of any
failures that would invalidate the above conclusions. 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 a margin of safety.
The proposed amendment extends the CHANNEL CALIBRATION
surveillance frequency for the ADS timers from 18 months to 24
months to facilitate a change in the DAEC operating cycle from 18
months to 24 months. Although the proposed change will result in an
increase in the interval between surveillance tests, the impact on
system availability is considered small based on other more frequent
testing, the availability of redundant systems or equipment, and the
fact that there is no evidence of any existing equipment failures
that would impact the availability of the ADS. Furthermore, an
evaluation of surveillance test results shows that the probability
of exceeding the TS AV with the extended surveillance frequency is
small and remains well within the setpoint methodology guideline.
Therefore, the proposed amendment will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
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: August 30, 2001.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) safety limit minimum critical
power ratio (MCPR) for two recirculation pump operation for Cycle 21.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed Safety Limit MCPR (SLMCPR), and its use to
determine the Cycle 21 thermal limits, have been derived using NRC
approved methods and uncertainties. These methods do not change
operation of the plant, and have no effect on the probability of an
accident initiating event or transient. The basis of the SLMCPR is
to ensure no mechanistic fuel damage is calculated to occur if the
limit is not violated. The new SLMCPR for Cycle 21 preserves the
margin to transition boiling and the probability of fuel damage is
not increased.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change results only from different inputs, for the
Cycle 21 core reload. These methods and uncertainties have been
reviewed and approved by the NRC, and do not involve any new or
unapproved methods for operating the facility. No new initiating
events or transients result from these changes.
The SLMCPR remains high enough to ensure that greater than 99.9%
of all fuel rods in the core will avoid transition boiling if the
limit is not violated, thereby preserving the fuel cladding
integrity. A change in SLMCPR cannot create the possibility of any
new type of accident. SLMCPR values for the new fuel cycle are
calculated using previously transmitted methodology.
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 amendment will not involve a significant
reduction in the margin of safety.
The margin of safety as defined in the TS bases will remain the
same. The new SLMCPR was derived using NRC approved methods and
uncertainties which are in accordance with the current fuel design
and licensing criteria. The SLMCPR remains high enough to ensure
that greater than 99.9% of all fuel rods in the core will avoid
transition boiling if the limit is not violated, thereby preserving
the fuel cladding integrity.
Fuel licensing acceptance criteria for SLMCPR calculations apply
to Monticello Cycle 21 in the same manner as previously applied.
SLMCPRs prepared using methodology previously transmitted to the NRC
ensure that greater than 99.9% of all fuel rods in the core will
avoid transition boiling if the limit is not violated, thereby
preserving fuel cladding integrity. The
[[Page 50471]]
operating MCPR limit is set appropriately above the safety limit
value to ensure adequate margin when the cycle specific transients
are evaluated.
Therefore, the proposed TS change does not involve a reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Claudia M. Craig.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: June 8, 2001.
Description of amendment request: The proposed amendment would
delete the ``High Pressure Coolant Injection (HPCI) System Suppression
Pool Water Level--High'' function from Technical Specification (TS)
3.3.5.1, ``Emergency Core Cooling System (ECCS) Instrumentation.'' This
change would eliminate automatic transfer of the HPCI pump suction
source from the condensate storage tank (CST) to the suppression pool
for a high suppression pool level. Elimination of this function is
expected to increase the availability of the HPCI system during a
postulated anticipated transient without scram (ATWS) with standby
liquid control system (SLCS) failure and to reduce operator burden
during a postulated station blackout (SBO) event.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
Deletion of the automatic HPCI suction transfer from the CST to
the suppression pool for a high suppression pool level condition was
analyzed for impacts against all previously evaluated accidents and
transients. Eliminating the automatic transfer increases the
availability of the HPCI system during an ATWS event and operator
burden is reduced during a postulated Station Blackout (SBO). There
are no adverse effects, consequences, or changes in the probability
of an accident occurring as a result of this change. HPCI operation
is improved and all other plant systems remain unaffected in their
ability to perform their design basis functions as a result of this
change.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously analyzed?
Implementation of this change increases the availability of the
HPCI system during a postulated ATWS with Standby Liquid Control
System (SLCS) failure. The change only affects the HPCI suction
source and whether the source is automatically transferred from the
preferred CST to the suppression pool for a high suppression pool
level. Continued HPCI operation utilizing the CST as a suction
source does not create a new or different type of accident from
those previously analyzed. The primary effect of this change is to
the suppression pool level which has been evaluated and found to be
acceptable for all relevant accidents and transients. Therefore a
new or different accident is not created and all other accident
analyses are unaffected by the change.
3. Does the proposed change involve a significant reduction in a
margin of safety.
This change does not reduce any margin of safety. The increase
in suppression pool water level does not cause containment
hydrodynamic loads to exceed design limits under accident
conditions. Overall, HPCI reliability is increased as it would
remain operable during the ATWS with Loss of SLCS event. This
increased availability of the HPCI system provides for additional
defense in depth which reduces the probability of core damage.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: Peter Tam, Acting.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: July 17, 2001.
Description of amendment request: The proposed amendment would
revise the reactor pressure vessel (RPV) pressure-temperature (P-T)
limits specified in the technical specifications (TSs). Editorial
changes associated with the P-T limit revisions are also proposed. The
proposed P-T limits rely on the methodology for determining allowable
P-T limits specified in American Society of Mechanical Engineers (ASME)
Code Case N-640. The revised P-T limits will allow required RPV
hydrostatic and leak tests to be performed at a significantly lower
temperature. This is expected to reduce challenges to plant operators
associated with maintaining the reactor coolant system within a narrow
temperature band during testing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated?
The changes to the calculational methodology for the pressure
and temperature (P-T) limits based upon Code Case N-640 continue to
provide adequate margin in the prevention of a brittle-type fracture
of the reactor pressure vessel (RPV). The code case was developed
based upon the knowledge gained through years of industry
experience. P-T curves developed using the allowances of Code Case
N-640 indeed yield more operating margin. However, the experience
gained in the areas of fracture toughness of materials and pre-
existing undetected defects show that some of the existing
assumptions used for the calculation of P-T limits are unnecessarily
conservative and unrealistic. Therefore, providing the allowances of
the subject Code Case in developing the P-T limit curves will
continue to provide adequate protection against nonductile-type
fractures of the RPV.
The evaluation for the Unit 1 and Unit 2 P-T limit curves for 32
EFPYs was performed using the approved methodologies of 10 CFR 50,
Appendix G. The curves generated from these methods ensure the P-T
limits will not be exceeded during any phase of reactor operation.
Resolution of the current industry issues related to fluence
calculation methodology requires PPL to limit applicability of the
curves to May 1, 2005 for Unit 2 and May 1, 2006 for Unit 1.
Therefore, the probability of occurrence and the consequences of a
previously analyzed event are not significantly increased. Finally,
the proposed changes will not affect any other system or piece of
equipment designed for the prevention or mitigation of previously
analyzed events. Thus, the probability of occurrence and the
consequences of any previously analyzed event are not significantly
increased as the result of the proposed changes.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously analyzed.
The proposed changes provide more operating margin in the P-T
limit curves for inservice leakage and hydrostatic pressure testing,
non-nuclear heatup and cooldown, and criticality, with the benefits
being primarily realizable during the pressure tests. Operation in
the ``new'' regions of the newly developed P-T curves has been
analyzed in accordance with the provisions of ASME Code, Section XI,
Appendix G; 10 CFR 50 Appendix G, and ASME Code Case N-640, thus
providing adequate protection against a
[[Page 50472]]
nonductile-type fracture of the RPV. These proposed changes do not
create the possibility of any new or different type of accident.
Further, they do not result in any new or unanalyzed operation of
any system or piece of equipment important to safety.
3. Does the proposed change involve a significant reduction in
the margin of safety?
As mentioned previously, the revised P-T curves provide more
operating margin and thus, more operational flexibility than the
current P-T curves. However, the industry experience since the
inception of the P-T limits in 1974 confirms that some of the
existing methodologies used to develop P-T curves is unrealistic and
unnecessarily conservative. Accordingly, ASME Code Case N-640 takes
advantage of the acquired knowledge by establishing more realistic
methodologies for the development of P-T curves.
Use of Code Case N-640 to develop the revised P-T curves
utilized the KIC fracture toughness curve in lieu of the
KIA curve as the lower bound for fracture toughness. Use
of the KIC curve to determine lower bound fracture
toughness is more technically correct than using the KIA
curve. P-T curves based on the KIC fracture toughness
limits enhance overall plant safety by expanding the P-T window in
the low-temperature operating region. The benefits which occur are a
reduction in the duration of the pressure test and personnel safety
while conducting inspections in primary containment with no decrease
to the margin of safety.
Therefore, operational flexibility is gained without a reduction
in the margin of safety to RPV brittle fracture.
The development of the P-T curves to 32 EFPYs was performed per
the guidelines of 10 CFR [part] 50, and thus, the margin of safety
is not reduced as the result of the proposed changes. Resolution of
the current industry issues related to fluence calculation
methodology requires PPL to limit applicability of the curves to May
1, 2005 for Unit 2 and May 1, 2006 for Unit 1.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, Inc., 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: Peter Tam, Acting.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: August 24, 2001.
Description of amendment requests: The licensee requests an
amendment to the Technical Specifications for containment leakage rate
testing as a result of recalculation of peak containment internal
pressure following certain design-basis accidents. The purpose of the
change is to make the Technical Specifications appropriately reflect
up-to-date calculated peak containment pressure. The revised calculated
peak containment pressure related to the design basis loss-of-coolant
accident and the revised calculated peak containment pressure for the
design basis Main Steam Line Break would be lower than the current
Technical Specification values.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the proposed amendment involve a significant increase
in the probability or consequences of an accident previously
evaluated?
Response: No.
The proposed change would revise the Operating Licenses for San
Onofre Nuclear Generating Station Units 2 and 3 to amend Technical
Specification (TS) 5.5.2.15, ``Containment Leakage Rate Testing
Program,'' by changing the stated calculated values for peak
containment internal pressure for the design basis Loss Of Coolant
Accident (LOCA) and Main Steam Line Break (MSLB) accident. The
current LOCA value of 55.1 psig would be changed to 45.9 psig and
the current MSLB value of 56.6 psig would be changed to 56.5 psig.
The proposed change does not affect the probability of
occurrence of an accident previously evaluated because it relates
solely to the consequences of hypothesized accidents given that the
accident has already occurred.
The proposed change does not increase the calculated peak
containment internal pressure for the LOCA and MSLB accidents, and
thus does not increase their consequences.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change relates to two accidents, MSLB and LOCA,
already evaluated in the Updated Final Safety Analysis Report.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
The recalculated peak containment internal pressures for the
MSLB and LOCA accidents are less than the containment design
pressure and less than the previously calculated pressures.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of amendment request: August 16, 2001.
Description of amendment request: The proposed amendments delete
requirements from the Technical Specifications (TS) (and, as
applicable, other elements of the licensing bases) to maintain a Post
Accident Sampling System (PASS). Licensees were generally required to
implement PASS upgrades as described in NUREG-0737, ``Clarification of
TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power
Plants to Assess Plant and Environs Conditions During and Following an
Accident.'' Implementation of these upgrades was an outcome of the
lessons learned from the accident that occurred at TMI, Unit 2.
Requirements related to PASS were imposed by Order for many facilities
and were added to or included in the TS for nuclear power reactors
currently licensed to operate. Lessons learned and improvements
implemented over the last 20 years have shown that the information
obtained from PASS can be readily obtained through other means or is of
little use in the assessment and mitigation of accident conditions.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 11, 2000 (65 FR 49271) on possible
amendments to eliminate PASS, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on October 31, 2000 (65 FR 65018). The licensee affirmed the
[[Page 50473]]
applicability of the following NSHC determination in its application
dated August 16, 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From any Previously Evaluated
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radionuclides
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested changes do not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment requests
involve no significant hazards consideration.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Section Chief: Richard L. Emch, Jr.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: August 2, 2001.
Description of amendment request: The proposed amendments delete
requirements from the Technical Specifications (TSs) (and, as
applicable, other elements of the licensing bases) to maintain a Post
Accident Sampling System (PASS). Licensees were generally required to
implement PASS upgrades as described in NUREG-0737, ``Clarification of
TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power
Plants to Assess Plant and Environs Conditions During and Following an
Accident.'' Implementation of these upgrades was an outcome of the
lessons learned from the accident that occurred at TMI, Unit 2.
Requirements related to PASS were imposed by Order for many facilities
and were added to or included in the TSs for nuclear power reactors
currently licensed to operate. Lessons learned and improvements
implemented over the last 20 years have shown that the information
obtained from PASS can be readily obtained through other means or is of
little use in the assessment and mitigation of accident conditions.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on August 11, 2000 (65 FR 49271) on possible
amendments to eliminate PASS, including a model safety evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on October 31, 2000 (65 FR 65018). The licensee affirmed the
applicability of the following NSHC determination in its application
dated August 2, 2001.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in
[[Page 50474]]
aiding the plant staff in assessing the extent of core damage and
subsequent offsite radiological dose projections. The system was not
intended to and does not serve a function for preventing accidents
and its elimination would not affect the probability of accidents
previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From any Previously Evaluated
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radionuclides
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested changes do not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment requests
involve no significant hazards consideration.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Robert A. Gramm.
Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of amendment request: January 18, 2001, as supplemented by
letter dated June 26, 2001.
Brief description of amendment request: The proposed amendment
would revise the applicability of the current BVPS-1 heatup/cooldown
curves from 15 effective full-power years (EFPY) to 14 EFPY. Proposed
changes to Technical Specification (TS) 3.7.1.1, ``Main Steam Safety
Valves (MSSVs),'' include revisions of the limiting condition for
operation and to the title and content of Table 3.7-1 to provide
consistency with the improved standard TSs, creation of new Actions to
address inoperable MSSVs, reduction of the power range neutron flux-
high reactor trip setpoint to be consistent with TS Traveler Form--235,
Revision 1, and changes to the maximum power levels permissible with
inoperable MSSVs. TS Bases changes are also proposed for consistency.
Date of publication of individual notice in Federal Register: July
27, 2001 (66 FR 39212).
Expiration date of individual notice: August 27, 2001.
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
[[Page 50475]]
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 e-mail to [email protected].
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: July 2, 2001.
Brief description of amendments: The amendments deleted Technical
Specifications Section 5.5.4, ``Post Accident Sampling,'' for Catawba
Nuclear Station, Units 1 and 2, and thereby eliminated the requirements
to have and maintain the post-accident sampling systems.
Date of issuance: September 11, 2001.
Effective date: As of the date of issuance and shall be implemented
within 180 days from the date of issuance.
Amendment Nos.: 193 and 185.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 8, 2001 (66 FR
41615).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 11, 2001.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: July 2, 2001.
Brief description of amendments: The amendments deleted Technical
Specifications (TS) Section 5.5.4, ``Post Accident Sampling,'' for
McGuire Nuclear Station, Units 1 and 2, and thereby eliminated the
requirements to have and maintain the post-accident sampling systems
(PASS). The amendments also delete PASS-related License Conditions
2.C(11)c, ``Post Accident Sampling (II.B.3),'' for Unit 1 and 2.C(10)b,
``Postaccident Sampling (II.B.3),'' for Unit 2.
Date of issuance: September 17, 2001.
Effective date: As of the date of issuance and shall be implemented
within 180 days from the date of issuance.
Amendment Nos.: 199 and 180.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 8, 2001 (66 FR
41616).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 17, 2001.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: February 28, 2001, supplemented
June 27, 2001.
Brief description of amendments: The amendments add new Technical
Specification 3.3.28 and Bases B 3.3.28 governing the addition of the
low pressure service water standby pump automatic start circuitry.
Date of Issuance: September 6, 2001.
Effective date: As of the date of issuance and shall be implemented
before the end of the Oconee Unit 3 End of Cycle 19 Refueling Outage.
Amendment Nos.: 319, 319, and 319.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 21, 2001 (66 FR
1517).
The supplement dated June 27, 2001, provided clarifying information
that did not change the scope of the February 28, 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 September 6, 2001.
No significant hazards consideration comments received: No.
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: January 24, 2001, as supplemented by
letter dated March 22, 2001.
Brief description of amendment: The amendment changes the limit on
the Low Power Setpoint, from 20 percent to 10 percent power, as
specified in Technical Specification (TS) 3.1.3, ``Control Rod
OPERABILITY,'' TS 3.1.6 ``Control Rod Pattern,'' and TS 3.3.2.1,
``Control Rod Block Instrumentation.''
Date of issuance: September 7, 2001.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 118.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 21, 2001 (66 FR
15921).
The supplemental letter dated March 22, 2001, provided additional
information that did not expand the scope of the application or change
the staff's initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 7, 2001.
No significant hazards consideration comments received: No.
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: January 24, 2001, as supplemented by
letters dated July 20 and August 7, 2001.
Brief description of amendment: The amendment request proposes
changes to the Technical Specifications (TSs) concerning certain
operational conditions required when conducting core alterations or
handling irradiated fuel in the primary containment. In addition, the
licensee proposes to implement administrative controls in accordance
with draft NUMARC 93-01, ``Industry Guidelines for Monitoring the
Effectiveness of Maintenance at Nuclear Power Plants,'' Revision 3,
Section 11.3.6.5, ``Containment--Primary (PWR [pressurized-water
reactor])/ Secondary (BWR [boiling-water reactor]),'' Revision 3,
Section 11.3.6, ``Assessment Methods for Shutdown Conditions,'' in lieu
of License Condition 2.C.(17) and change terms to make them consistent
with the terminology in other revised TSs.
Date of issuance: September 14, 2001.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 119.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: April 18, 2001 (66 FR
20001).
The supplemental letters dated July 20 and August 7, 2001, provided
[[Page 50476]]
additional information that did not expand the scope of the application
or change the staff's initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 14, 2001.
No significant hazards consideration comments received: No.
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: January 24, 2001, as supplemented by
letters dated July 2, and August 6 and 20, 2001.
Brief description of amendment: The license amendment request
consists of changes to the Technical Specifications (TSs) to revise the
reactor vessel pressure/temperature (P/T or P-T) limits specified in TS
3.4.11, ``RCS [Reactor Coolant System] Pressure and Temperature (P/T)
Limits,'' for reactor heat-up. The current RCS P/T Limits in TS Figure
3.4-11, ``Minimum Temperature Required Vs. RCS Pressure,'' would be
replaced with recalculated RCS P/T limits based, in part, on an
alternate methodology. The alternate methodology uses American Society
of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code
(Code) Case N-640, ``Alternative Requirement Fracture Toughness for
Development of P-T Limit Curves for ASME B&PV Code Section XI, Division
1,'' for alternate reference fracture toughness for reactor vessel
materials in determining the P/T limits.
Date of issuance: September 14, 2001.
Effective date: As of the date of issuance and shall be implemented
30 days from the date of issuance.
Amendment No.: 120.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 21, 2001 (66 FR
15920).
The supplemental letters dated July 2, and August 6 and 20, 2001,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 14, 2001.
No significant hazards consideration comments received: No.
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: April 27, 2001.
Brief description of amendments: The amendments revise Technical
Specifications (TS) Section 5.5.7, to provide an exception to the
recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14,
Revision 1, which would allow either a qualified in-place ultrasonic
volumetric examination over the volume from the inner bore of the
flywheel to the circle of one-half the outer radius or a surface
examination (magnetic particle testing and/or liquid penetrant testing)
of exposed surfaces of the removed flywheel to be conducted at
approximately 10-year intervals. The proposed change is in accordance
with the Nuclear Regulatory Commission (NRC) approved Improved Standard
TS Generic Change Traveler TSTF-237, Revision 1.
Date of issuance: September 20, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 123, 123, 118, and 118.
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: The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated September 20, 2001.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: June 26, 2001.
Brief description of amendments: To extend the dates specified in
Operating License Sections 2.C(8) and 3.P, ``Pressure--Temperature
Limit Curves,'' for Dresden Nuclear Power Station, Units 2 and 3,
respectively.
Date of issuance: September 10, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 187 and 182.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Facility Operating License.
Date of initial notice in Federal Register: The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated September 10, 2001.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: March 21, 2001, as supplemented
June 28, 2001.
Brief description of amendment: The amendment revised the Improved
Technical Specifications (ITS) 5.6.2.10, ``OTSG [Once-Through Steam
Generator] Tube Surveillance Program'' to implement a reroll process to
repair degraded steam generator tubes and allow the reroll repairs to
be used in both the upper and lower tubesheets.
Date of issuance: September 10, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 198.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 18, 2001 (66 FR
20006). The supplemental letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 10, 2001.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: October 3, 2000, as supplemented
June 14, August 28, and September 7, 2001.
Brief description of amendment: The amendment revised the Improved
Technical Specifications (ITS) 3.7.12, ``Control Room Emergency
Ventilation System (CREVS)''; ITS 5.6.2.12, ``Ventilation Filter
Testing Program (VFTP)''; ITS 3.3.16, ``Control Room Isolation--High
Radiation''; and ITS 3.7.18, ``Control Complex Cooling System.'' The
proposed ITS changes are based on the results of revised public and
control room dose calculations for CR-3 design basis radiological
accidents using an alternative source term and the adoption of
Technical Task Force Traveler (TSTF) 287. A new Section 5.6.2.21,
``Control Complex Habitability Envelope Program,'' is added.
[[Page 50477]]
Date of issuance: September 17, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 199.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 15, 2000 (65
FR 69060). The supplemental letters provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 17, 2001.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of application for amendment: April 18, 2001, as supplemented
August 24, 2001.
Brief description of amendment: This amendment revises Technical
Specification (TS) 6.9.1.11, ``Core Operating Limits Report (COLR),''
to include the ABB-Combustion Engineering Topical Report CENPD-387-P-A,
Rev 000, in the list of analytical methods. This allows use of an
improved heat flux correlation (designated ABB-NV) previously approved
by the NRC. Additionally, the Bases for TS 2.1.1, ``Reactor Core,'' are
modified to reflect use of the improved heat flux correlation.
Date of Issuance: September 20, 2001.
Effective Date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 118.
Facility Operating License No. NPF-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 30, 2001 (66 FR
29358). The August 24, 2001, supplement did not affect the original
proposed no significant hazards determination, or expand the scope of
the request as noticed in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 20, 2001.
No significant hazards consideration comments received: No.
GPU Nuclear Inc., Docket No. 50-320, Three Mile Island Nuclear
Generating Station, Unit 2, Dauphin County, Pennsylvania
Date of amendment request: July 25, 2000, as supplemented by letter
dated June 21, 2001.
Brief description of amendment request: The amendment changes Three
Mile Island Nuclear Generating Station, Unit 2 (TMI-2), Technical
Specification (TS) 6.7.2 to eliminate a change associated with periodic
reviews of procedures. Currently, TS 6.7.2 states that required
procedures shall be reviewed periodically as required by American
National Standards Institute (ANSI) Standard N18.7-1976 (a biennial
review). This amendment revises the wording for TS 6.7.2 to state that
required procedures shall be reviewed periodically. This amendment is
also consistent with the TMI-2 Post-Defueling Monitored Storage Quality
Assurance Plan, which states that ``Procedural documentation shall be
periodically reviewed for adequacy as set forth in administrative
procedures.''
Date of Issuance: September 7, 2001.
Effective Date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 56.
Facility Operating License No. DPR-73: Amendment revises the
Technical Specification.
Date of initial notice in Federal Register: December 13, 2000 (66
FR 77920).
The June 21, 2001, supplemental letter replaced in its entirety the
original application dated July 25, 2000. The supplement did not expand
the scope of the original request.
The Commission's related evaluation is contained in a safety
evaluation dated September 7, 2001.
No significant hazards consideration comments received: No.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: August 17, 2001.
Brief description of amendment: The amendment allowed a one-time
exception to Technical Specification Surveillance Requirement (SR)
3.6.1.7.2 for suppression chamber-to-drywell vacuum breakers 2ISC*RV35A
and 2ISC*RV35B. A note has been added to SR 3.6.1.7.2 stating that
function testing of these vacuum breakers is not required to be met for
the remainder of Cycle 8.
Date of issuance: September 7, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 98.
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Public Comments Requested as to Proposed No Significant Hazards
Consideration: Yes (66 FR 44653) August 24, 2001. That notice provided
an opportunity to submit comments on the Commission's proposed no
significant hazards consideration determination. Comments were received
from one person, and were addressed in the safety evaluation associated
with the amendment. The notice also provided for an opportunity to
request a hearing by September 24, 2001, but indicated that if the
Commission makes a final no significant hazards consideration
determination, any such hearing would take place after the issuance of
the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, final determination of no significant hazards
consideration determination, and state consultation, are contained in a
safety evaluation dated September 7, 2001.
Attorney for the Licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Peter S. Tam, Acting.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of application for amendment: April 6, 2001.
Brief description of amendment: The amendment changes Technical
Specifications Section 5.5.10, ``Technical Specifications (TS) Bases
Control Program,'' in accordance with Nuclear Energy Institute TS Task
Force Standard TS Change Traveler, TSTF-364, ``Revision to TS Bases
Control Program to Incorporate Changes to 10 CFR 50.59,'' Revision 0.
Date of issuance: September 13, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 241.
Facility Operating License No. DPR-49: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 8, 2001 (66 FR
41623).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 13, 2001.
No significant hazards consideration comments received: No.
[[Page 50478]]
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: May 25, 2001, as supplemented
August 17, 2001.
Brief description of amendment: The amendment to the Kewaunee
Nuclear Power Plant Technical Specifications (TSs) 4.2 revises TS 4.2
to revise the surveillance requirements and bases for TS 4.2.b, ``Steam
Generator Tubes,'' to account for changes associated with replacement
of the original steam generators. Specifically, the changes delete
inspection requirements associated with steam generator tube sleeving
and repair limits and revise the phrasing of text within the TS to
enhance clarity.
Date of issuance: September 20, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 158.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 12, 2001 (66 FR
31711).
The supplemental information contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice. The Commission's related evaluation of the amendment
is contained in a Safety Evaluation dated September 20, 2001.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: May 24, 2001.
Brief description of amendments: The amendments revise the
Technical Specification definition of CORE ALTERATIONS.
Date of issuance: September 11, 2001.
Effective date: The amendments are effective as of the date of
their issuance and shall be implemented within 30 days from the date of
issuance.
Amendment Nos.: Unit 1--131; Unit 2--120.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 11, 2001 (66 FR
36345)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 11, 2001.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendments request: May 24, 2001.
Brief description of amendments: The amendments relocate Technical
Specification 3/4.9.6, ``Refueling Machine'' and its associated Bases
description to the Technical Requirements Manual.
Date of issuance: September 13, 2001.
Effective date: The amendments are effective as of the date of
their issuance.
Amendment Nos.: Unit 1--132; Unit 2--121.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 11, 2001 (66 FR
36344).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 13, 2001.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-296, Browns Ferry Nuclear
Plant, Units 3 Limestone County, Alabama
Date of application for amendment: July 25, 2001.
Brief description of amendment: The proposed amendment deletes
Technical Specification (TS)-required Action 3.3.1.1.I.2, which limits
plant operation to 120 days in the event of the inoperability of the
Oscillation Power Range Monitor trip system. For this situation, the
proposed change would allow plant operation to continue if the existing
TS Required Action 3.3.1.1.I.1, to implement an alternate means to
detect and suppress thermal hydraulic instability oscillations, was
taken.
Date of issuance: September 13, 2001.
Effective date: Date of issuance and shall be implemented within 30
days.
Amendment No.: 231.
Facility Operating License No. DPR-68: Amendment revises the TS.
Date of initial notice in Federal Register: August 8, 2001 (66 FR
41627).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 13, 2001.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: June 16, 2000, as supplemented
September 27, 2000, and June 6, 2001.
Brief Description of amendments: These amendments change the
reactor protection system and engineered safety features actuation
system analog instrumentation surveillance frequency from monthly to
quarterly.
Date of issuance: August 31, 2001.
Effective date: August 31, 2001.
Amendment Nos.: 228 and 228.
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: November 15, 2000 (65
FR 69067). The September 27, 2000, and June 6, 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 August 31, 2001.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
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 for the
amendment 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.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal
[[Page 50479]]
Register notice providing opportunity for public comment or has used
local media to provide notice to the public in the area surrounding a
licensee's facility of the licensee's application and of the
Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective 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
application for amendment, (2) the amendment to Facility Operating
License, 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 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].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By November 2, 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, 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. Since the Commission has made a final determination
that the
[[Page 50480]]
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
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 20555-001, Attention: Rulemakings and
Adjudications Staff, 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 20555-001, 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 the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Oswego County, New York
Date of amendment request: September 14, 2001.
Brief description of amendment: The amendment authorizes a one-
time-only change to Technical Specifications Section 3.9.B.1 and
associated Bases. Specifically, this change extends the Limiting
Condition for Operation allowable out-of-service time for one incoming
Reserve AC Power line (115KV line #3) and/or one reserve station
transformer inoperable from 7 days to 14 days during the period
commencing September 9, 2001 and extending through September 23, 2001.
Date of issuance: September 15, 2001.
Effective date: As of the date of issuance.
Amendment No.: 272.
Facility Operating License No. DPR-59: Amendment revises the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: No.
The Commission's related evaluation of the amendment, finding of
emergency circumstances, and final determination of no significant
hazards consideration, are contained in a Safety Evaluation dated
September 14, 2001.
Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Section Chief: Peter Tam (Acting)
Note:
The publication date for this notice will change from every
other Wednesday to every other Tuesday, effective January 8, 2002.
The notice will contain the same information and will continue to be
published biweekly.
Dated at Rockville, Maryland, this 25th day of September 2001.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 01-24580 Filed 10-2-01; 8:45 am]
BILLING CODE 7590-01-P