[Federal Register Volume 68, Number 121 (Tuesday, June 24, 2003)]
[Notices]
[Pages 37574-37590]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-15597]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from, May 30, 2003, through June 12, 2003. The
last biweekly notice was published on June 10, 2003 (68 FR 28844).
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
[[Page 37575]]
Federal Register a notice of issuance and provide for opportunity for a
hearing after issuance. The Commission expects that the need to take
this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By July 24, 2003, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714, which is
available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records will be accessible from the
Agencywide Documents Access and Management System's (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing
or petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of continuing disruptions in delivery of mail to United States
Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to (301) 415-
1101 or by e-mail to [email protected]. A copy of the request for
hearing and petition for leave to intervene should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and because of continuing disruptions in
delivery of mail to United States Government offices, it is requested
that copies be transmitted either by means of facsimile transmission to
(301) 415-3725 or by e-mail to [email protected]. A copy of the
request for hearing and petition for leave to intervene should also be
sent to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's
[[Page 37576]]
PDR, located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland. Publicly available
records will be accessible from the Agencywide Documents Access and
Management System's (ADAMS) Public Electronic Reading Room on the
Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html.
If you do not have access to ADAMS or if there are problems in
accessing the documents located in ADAMS, contact the NRC PDR Reference
staff at 1-800-397-4209, (301) 415-4737 or by e-mail to [email protected].
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: May 12, 2003.
Description of amendments request: The proposed amendment would
extend several Required Action Completion times for inoperable diesel
generators (DGs) identified in Technical Specification 3.8.1, ``AC
Sources-Operating.''
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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed Technical Specification changes do not affect the
design, operational characteristics, function or reliability of the
DGs. The DGs are not accident initiators, and extending the DG
Required Action Completion Times will not impact the frequency of
any previously evaluated accidents. The design basis accidents will
remain the same postulated events described in the Updated Final
Safety Analysis Report. In addition, extending the DG Required
Action Completion Times will not impact the consequences of an
accident previously evaluated. The consequences of previously
evaluated accidents will remain the same during the proposed
extended Required Action Completion Times as during the current
Required Action Completion Times. The ability of the remaining DGs
to mitigate the consequences of an accident will not be affected
since no additional failures are postulated while equipment is
inoperable within the Technical Specification Required Action
Completion Times. Therefore, the proposed changes will not increase
the probability or consequences of an accident previously evaluated.
The duration of a Technical Specification Required Action
Completion Time is determined considering that there is a minimal
possibility that an accident will occur while a component is removed
from service. A risk informed assessment was performed that
concluded that the plant risk is acceptable and consistent with the
guidance contained in Regulatory Guide 1.177.
The additional proposed changes to renumber action requirements
and the correction of a misspelled word will not result in any
technical changes to the current requirements. Therefore, these
additional proposed changes will not increase the probability or
consequences of an accident previously evaluated.
2. Would not create the possibility of a new or different [kind]
of accident from any accident previously evaluated.
The proposed changes to the Technical Specifications do not
impact any system or component in a manner that could cause an
accident. The proposed changes will not alter the plant
configuration or require any unusual operator actions. The proposed
changes will not alter the way any structure, system, or component
functions, and will not significantly alter the manner in which the
plant is operated. There will be no adverse effect on plant
operation or accident mitigation equipment. The response of the
plant and the operator following an accident will not be
significantly different. In addition, the proposed changes do not
introduce any new failure modes. Therefore, the proposed changes
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The margin of safety provided by the DGs is to provide emergency
back-up power supply to systems required to mitigate the
consequences of postulated accidents. The engineered safety features
systems on either of the two trains for each unit provide for the
minimum safety functions necessary to shutdown the units and
maintain it in a safe shutdown condition. Each of the two trains can
be powered from one of the offsite power sources or its associated
DG. In addition, the 0C DG (Station Blackout DG) is available to
provide power to any of the trains. This design provides adequate
defense in-depth to ensure that diverse power sources are available
to accomplish the required safety functions. Thus, with a safety-
related DG out-of-service, there is sufficient means to accomplish
the safety functions and prevent the release of radioactive material
in the event of an accident.
The proposed change does not affect any of the assumptions or
inputs to the Updated Final Safety Analysis Report and does not
reduce the decrease in severe accident risk achieved with the
issuance of the Station Blackout Rule, 10 CFR 50.63, ``Loss of All
Alternating Current Power.''
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
amendments request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Richard J. Laufer.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: February 26, 2003.
Description of amendment request: The proposed amendments would
allow the licensee to revise the Updated Final Safety Analysis Report
to include a description of a load drop analysis performed for handling
reactor cavity shield blocks weighing greater than 110 tons with the
Dresden, Units 2 and 3, reactor building crane.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change will allow use of a load drop analysis
performed for handling the reactor cavity shield blocks weighing
greater than 110 tons with the reactor building crane during power
operation. The load drop analysis demonstrates that dropping a
reactor cavity shield block within the designated safe load path
from the heights assumed in the analysis will not affect the
capability of safety-related equipment to perform its function.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change will allow use of a load drop analysis
performed for handling the reactor cavity shield blocks weighing
greater than 110 tons with the reactor building crane during power
operation. The load drop analysis demonstrates that dropping a
reactor cavity shield block within the designated safe load path
from the heights assumed in the analysis will not affect the
capability of safety-related equipment to perform its function.
Therefore, the proposed change will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed change will allow use of a load drop analysis
performed for handling the reactor cavity shield blocks weighing
[[Page 37577]]
greater than 110 tons with the reactor building crane during power
operation. The load drop analysis demonstrates that dropping a
reactor cavity shield block within the designated safe load path
from the heights assumed in the analysis will not affect the
capability of safety-related equipment to perform its function.
Therefore, it is concluded that the proposed change does not result
in a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois.
Date of amendment request: May 19, 2003.
Description of amendment request: The proposed amendments would
revise Appendix A, Technical Specifications (TS), of Facility Operating
License Nos. NPF-11 and NPF-18. Specifically, the proposed change will
decrease the frequency associated with TS Surveillance Requirement (SR)
3.7.7.1 for Turbine Bypass Valve (BPV) testing from 7 to 31 days. The
proposed change is consistent with the testing frequency contained in
NUREG-1434, ``Standard Technical Specifications General Electric
Plants, BWR/6,'' Revision 2, dated June 2001, for BPV testing.
The 7-day frequency associated with SR 3.7.7.1 was established in
the LaSalle County Station (LSCS) TS during conversion to Improved
Technical Specifications (ITS) format due to the testing frequency
contained in the LSCS custom TS and the difficulties experienced with
other Electro-Hydraulic Control (EHC) system valves to consistently
pass their surveillance tests. LSCS has recently re-evaluated the
performance of these valves and has determined that the current
performance of these valves supports decreasing the testing frequency
of the BPVs from 7 to 31 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in probability or consequences of an accident previously evaluated.
The proposed change will decrease the frequency associated with
Surveillance Requirement (SR) 3.7.7.1 for turbine bypass valve (BPV)
testing from 7 to 31 days. The proposed change is consistent with
the testing frequency contained in NUREG-1434, ``Standard Technical
Specifications General Electric Plants, BWR/6,'' Revision 2, dated
June 2001, for BPV testing. The performance of BPV surveillance
testing is not a precursor to any accident previously evaluated.
Thus, the proposed change does not have any effect on the
probability of an accident previously evaluated.
The Main Turbine Bypass System is required to be operable to
limit peak pressure in the main steam lines and maintain reactor
pressure within acceptable limits during events that cause rapid
pressurization, such that the Safety Limit Minimum Critical Power
Ratio (MCPR) is not exceeded. An operable Main Turbine Bypass System
requires the BPVs to open in response to increasing main steam line
pressure. The performance of BPVs surveillance testing provides
assurance that the valves will operate as assumed in accidents
previously evaluated. Thus, the radiological consequences of any
accident previously evaluated are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not affect the control parameters
governing unit operation and does not introduce any new equipment,
modes of system operation or failure mechanisms. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed change will decrease the frequency associated with
SR 3.7.7.1 for BPV testing from 7 to 31 days. The proposed change is
consistent with the BPV testing frequency contained in NUREG-1433,
Revision 2, and does not affect the design parameters or the
setpoints associated with BPV operation. Therefore, the proposed
changes do not involve a significant reduction in a margin of
safety.
Based upon the above, Exelon Generation Company concludes that
the proposed amendment presents no significant hazards consideration
under the standards set forth in 10 CFR 50.92(c), and, accordingly,
a finding of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County,
Pennsylvania
Date of amendment request: March 26, 2003.
Description of amendment request: The proposed amendments would
modify Technical Specifications (TSs) 4.0.1 and 4.0.3 to be consistent
with the Improved Standard Technical Specifications. The proposed
amendments would also modify the TS requirements for missed
surveillances in TS 4.0.3 to be consistent with the Nuclear Regulatory
Commission (NRC)-approved Technical Specification Task Force (TSTF),
Standard Technical Specification Change TSTF-358, Revision 6.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments
concerning missed surveillances, including a model safety evaluation
and model no significant hazards consideration (NSHC) determination,
using the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on September 28, 2001 (66 FR 49714). The licensee affirmed the
applicability of that portion of the following NSHC determination,
related to the adoption of the TSTF-358, Revision 6, changes to the TSs
in its application dated March 26, 2003.
Basis for proposed no significant hazards consideration
determination:
Item 1: Modification of TSs 4.0.1 and 4.0.3 to be consistent with
the Improved Standard Technical Specifications.
As required by 10 CFR 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration, which is
presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes involve rewording of the existing Technical
Specifications to be
[[Page 37578]]
consistent with NUREG-1431, Revision 2. These modifications involve
no technical changes to the existing Technical Specifications. As
such, these changes are administrative in nature and do not affect
initiators of analyzed events or assumed mitigation of accident or
transient events. Therefore, the proposed changes will not increase
the probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes involve rewording of the existing Technical
Specifications to be consistent with NUREG-1431, Revision 2. The
change does not involve a physical alteration of the plant (no new
or different type of equipment will be installed) or changes in
methods governing normal plant operation. The changes will not
impose any new or different requirements or eliminate any existing
requirements. Therefore, the proposed changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes involve rewording of the existing Technical
Specifications to be consistent with NUREG-1431, Revision 2. The
changes are administrative in nature and will not involve any
technical changes. The changes will not reduce a margin of safety
because they have no impact on any safety analysis assumptions.
Also, since these changes are administrative in nature, no question
of safety is involved. Therefore, there will be no reduction in a
margin of safety.
Item 2: Incorporation of TSTF-358--Revision 6.
As required by 10 CFR 50.91(a), an analysis of the issue of no
significant hazards on consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated
The proposed change relaxes the time allowed to perform a missed
surveillance. The time between surveillances is not an initiator of
any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The
equipment being tested is still required to be operable and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected. Any reduction
in confidence that a standby system might fail to perform its safety
function due to a missed surveillance is small and would not, in the
absence of other unrelated failures, lead to an increase in
consequences beyond those estimated by existing analyses. The
addition of a requirement to assess and manage the risk introduced
by the missed surveillance will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of the
plant (no new or different type of equipment will be installed) or a
change in the methods governing normal plant operation. A missed
surveillance will not, in and of itself, introduce new failure modes
or effects and any increased chance that a standby system might fail
to perform its safety function due to a missed surveillance would
not, in the absence of other unrelated failures, lead to an accident
beyond those previously evaluated. The addition of a requirement to
assess and manage the risk introduced by the missed surveillance
will further minimize possible concerns. Thus, this change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The extended time allowed to perform a missed surveillance does
not result in a significant reduction in the margin of safety. As
supported by the historical data, the likely outcome of any
surveillance is verification that the LCO [Limiting Condition for
Operation] is met. Failure to perform a surveillance within the
prescribed frequency does not cause equipment to become inoperable.
The only effect of the additional time allowed to perform a missed
surveillance on the margin of safety is the extension of the time
until inoperable equipment is discovered to be inoperable by the
missed surveillance. However, given the rare occurrence of
inoperable equipment, and the rare occurrence of a missed
surveillance, a missed surveillance on inoperable equipment would be
very unlikely. This must be balanced against the real risk of
manipulating the plant equipment or condition to perform the missed
surveillance. In addition, parallel trains and alternate equipment
are typically available to perform the safety function of the
equipment not tested. Thus, there is confidence that the equipment
can perform its assumed safety function.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis of Item 1 and
the licensee's reference to the analysis included in the
consolidated line-item improvement process Federal Register Notice,
June 14, 2001 (66 FR 32400) for Item 2, 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: Richard J. Laufer.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-
Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: November 30, 2001.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) by decreasing the
pressurizer high level limit and by revising the required action
when the pressurizer is inoperable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The new pressurizer high level limit is more restrictive than
the existing limit, and accident initial conditions, probability,
and assumptions remain as previously analyzed. The proposed change
to the pressurizer allowed outage time will have no significant
effect on accident initiation frequency. The proposed changes do not
invalidate the assumptions used in evaluating the radiological
consequences of any accident. Therefore, the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not introduce any new or different
accident initiators. Therefore, the proposed changes do not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change to the pressurizer high level limit will
ensure an adequate margin of safety is maintained. The proposed
change to the pressurizer allowed outage time is minimal and will
not have a significant effect on any margin of safety. Therefore,
the proposed changes do not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
[[Page 37579]]
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: October 8, 2002.
Description of amendment request: Omaha Public Power District
(OPPD) has proposed the following changes to the Technical
Specifications (TSs): (1) Use a pressure temperature limits report
(PTLR), (2) change the minimum boltup temperature, (3) modify the TSs
to reflect the revised low temperature overpressure protection (LTOP)
methodology and analysis that is submitted for review and approval, (4)
perform LTOP analyses ``in-house,'' (5) change the LTOP enable
temperature, (6) modify TS 2.10.1 to exactly specify the reactor
coolant system (RCS) temperature at which the reactor can be made
critical, and (7) add a TS for a maximum pressure value for the safety
injection tanks. The use of a PTLR requires the relocation of TS Figure
2-1 (RCS Pressure--Temperature Limits for Heatup, Cooldown, and In-
service Test) into Figure 5-1 of the PTLR. As a result of these
changes, the following TSs are required to either be modified or added:
define the PTLR in Definitions; TS 2.1.1(8); TS 2.1.1(11); Basis
Section of TS 2.1.1; TS 2.1.2, including the TS 2.1.2 Basis and
Reference Sections; TS 2.1.6(4); TS 2.3(1)(c); TS 2.3(3); TS 2.3
References; TS 2.10.1 and TS 2.10.1 Basis Section; Table 3-5, item 23,
TS 3.3(1)(c); and TS 5.9.6.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes will not increase the probability or
consequence of any accident for the following reasons:
(1) The proposed changes relocate the Pressure--Temperature (P-
T) limit curves and low temperature over pressure protection (LTOP)
system setpoints to the Pressure and Temperature Limits Report
(PTLR). Compliance with these curves and limits continues to be
required by the Technical Specifications (TSs). Changes to the
curves will be controlled by TS 5.9.6, which contains the NRC
approved methodologies used in the development of the PTLR. The
change to the P-T limit curve as shown on Figure 5-1 of the PTLR is
in compliance with Reference 10.11 [of the licensee's October 8,
2002, submittal], Westinghouse Electric Company/Combustion
Engineering's (W/CE's) methodology and ASME Code Case N-640 for
performing P-T limit curves.
(2) Revisions to the LTOP system limits can only be made in
accordance with the approved methodologies stated in TS 5.9.6 with
any resulting setpoint changes controlled by the 10 CFR 50.59
process. The PTLR in combination with the limitations imposed by the
TSs will ensure the integrity of the reactor vessel pressure
boundary.
(3) The conservative, but lower minimum boltup temperature and
LTOP enable temperature are in compliance with Reference 10.12 [of
the licensee's October 8, 2002, submittal]. Since the P-T limit
curves and LTOP analysis are analyzed to the same temperatures as
these proposed temperature values, there is no reduction to the
margin of safety.
(4) Restricting the RCS temperature at which the reactor can be
made critical is more conservative than the minimum temperature
requirements for core critical operations based on fracture
mechanics considerations as required by Reference 10.11 [of the
licensee's October 8, 2002, submittal] during physics testing.
(5) Addition of a maximum pressure to the safety injection tanks
(SITs) ensures compliance with Criterion 2 of 10 CFR
50.36(c)(2)(ii).
Therefore, the probability or consequence of any accident is not
increased.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed revision does not change any equipment required to
mitigate the consequences of an accident. The continued use of the
same TS administrative controls prevents the possibility of a new or
different kind of accident. Since the proposed changes do not
involve the addition or modification of equipment nor alter the
design of plant systems, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated. The changes proposed do not change how design
basis accident events are postulated nor do the changes themselves
initiate a new kind of accident or failure mode with a unique set of
conditions (proposed administrative controls). Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
Relocating the P-T limit curves and LTOP system setpoints to the
PTLR is in compliance with Reference 10.7 [of the licensee's October
8, 2002, submittal]. Future updates of the PTLR will be conducted
under the 10 CFR 50.59 process utilizing NRC approved methodologies.
Updating the P-T limit curve is in accordance with Reference 10.11
[of the licensee's October 8, 2002, submittal], W/CE's methodology
and ASME Code Case N-640. Reduction of the minimum boltup
temperature and LTOP enable temperature is in compliance with
Reference 10.12 [of the licensee's October 8, 2002, submittal].
Restricting the reactor coolant system (RCS) temperature at which
the reactor can be made critical is more conservative than the
minimum temperature requirements for core critical operations based
on fracture mechanics considerations as required by Reference 10.11
[of the licensee's October 8, 2002, submittal], during physics
testing. Addition of a maximum pressure to the SITs is in accordance
with Criterion 2 of 10 CFR 50.36(c)(2)(ii). Additionally, the LTOP
methodology and analysis conforms to Reference 10.10 [of the
licensee's October 8, 2002, submittal]. Therefore, the proposed
changes do not involve a significant reduction to 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: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: April 2, 2003.
Description of amendment requests: The proposed license amendments
would revise Technical Specification (TS) 5.5.11, ``Ventilation Filter
Testing Program (VFTP),'' to change the surveillance frequency,
penetration, and relative humidity requirements for laboratory testing
of the charcoal adsorber for the control room, auxiliary building, and
fuel handling building ventilation systems. This would also eliminate
the charcoal preheater testing requirements. TS 3.7.10, ``Control Room
Ventilation System (CRVS),'' and TS 3.7.12, ``Auxiliary Building
Ventilation System (ABVS),'' will also be revised to be consistent with
these changes. These changes are in accordance with Regulatory Guide
1.52, Revision 3, ``Design, Inspection, and Testing Criteria for Air
Filtration and Adsorption Units of Post Accident Engineered-Safety-
Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power
Plants,'' Generic Letter 99-02, ``Laboratory Testing of Nuclear-Grade
Activated Charcoal,'' and the requirements in American Society for
Testing and Materials D3803-1989, ``Standard Technical Method for
Nuclear-Grade Activated Carbon.'' In addition, TS 3.7.10 would be
revised by adding a note allowing the control room boundary to be open
intermittently under administrative control; adding a new required TS
Action for two CRVS
[[Page 37580]]
trains being inoperable due to an inoperable control room boundary, and
revising the relettered Condition F to add ``for reasons other than
Condition B.'' TS Surveillance Requirement (SR) 3.7.12.3 would be
revised to limit its applicability and TS 3.7.13, ``Fuel Handling
Building Ventilation System (FHBVS),'' would be revised to add the word
``recently'' to qualify the irradiated fuel in the statement of
applicability. These proposed revisions are made consistent with NUREG-
1431, Revision 2, ``Standard Technical Specifications Westinghouse
Plant,'' April 2001, and limit unnecessary surveillance testing when
the ABVS is actively performing its safety function.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes revise the frequency (from 18 months to 24
months), and acceptance criteria for laboratory testing of the
charcoal adsorbers in the engineered safety feature (ESF)
ventilation systems. The testing is performed offsite on charcoal
samples taken from the ventilation systems, and would have no impact
on any accident initiator, or change the consequences of any
previously analyzed accident. Continued compliance with industry
standards and Diablo Canyon Power Plant test data ensure that the
revised requirements would continue to ensure the charcoal adsorbers
are capable of performing their intended safety function; therefore,
the changes would not affect the accident mitigation capabilities of
the ESF ventilation systems.
The preheaters in the control room ventilation system (CRVS) and
auxiliary building ventilation system (ABVS) are not initiators of
analyzed events, are no longer credited in mitigating design basis
accidents or transients, and are therefore not required for system
operability. The deletion of the requirement to demonstrate the
capability of the preheaters every 24 months, and the changes to the
action requirements and surveillance requirements for the CRVS and
ABVS would not affect the assumed accident mitigation capabilities
of these ESF ventilation systems.
The proposed changes also provide for two trains of the CRVS to
be inoperable for up to 24 hours as a result of the CRVS boundary
being inoperable. This allowance is contingent on providing and
implementing proceduralized compensatory measures to restore the
boundary during that time period. Although this change does provide
for an increase in the allowed time for continued plant operation in
the applicable modes, its acceptability is based on the low
probability of any design basis accident during that time period and
the protection provided by the compensatory measures that would be
established. In addition, this change has no impact on any accident
initiator, and does not change the consequences of any previously
analyzed accident, because the administrative controls will restore
the boundary before it is required to protect control room
personnel.
The proposed changes also provide for limiting the applicability
of surveillance requirement (SR) 3.7.12.3, which verifies the
operability of the ABVS on a safety injection (SI) signal. The
limitation is imposed only when the ABVS is aligned and operating in
its safety function configuration. Since the ABVS is already
performing its safety function when it is in that condition,
verifying the automatic capability to transfer to that configuration
is unnecessary. Since this limitation is only during periods where
the ABVS is in its safety function configuration it has no impact on
any accident initiator, or change the consequences of any previously
analyzed accident. In addition, this surveillance is still required
to be current whenever the ABVS is returned to automatic.
The proposed changes also provide for limiting the required
operability of the fuel handling building ventilation system (FHBVS)
based on a minimum time period that all fuel assemblies in the fuel
pool have not been part of a critical core. This change does reduce
the current operability requirements for the FHBVS and increases the
consequences of a fuel handling accident with the FHBVS inoperable.
However, limiting the FHBVS operability requirements does not
increase the probability of any accident, and as determined in the
new fuel handling accident (FHA) analysis, the potential release
levels are still well within acceptable limits and do not
significantly increase the consequences of a FHA.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The ABVS, FHBVS and CRVS are accident response systems and as
such cannot create accidents. The changes to the charcoal sample
test requirements will not affect the method of operation of the
systems. The proposed changes only affect the laboratory test
acceptance criteria for the charcoal samples, and how the charcoal
preheaters are credited for meeting technical specification (TS)
requirements. These changes result in a more conservative testing
methodology. Deletion of the preheater requirements from the TS is
based on the heaters not being credited for mitigation of any
accident condition and does not affect the operation of these
systems. The design and operation of the CRVS, ABVS, and FHBVS are
not affected by these changes. No new or different accident
scenarios, transient precursors, failure mechanisms, or limiting
single failures will be introduced as a result of these changes.
The proposed changes also provide for two trains of the CRVS to
be inoperable for up to 24 hours as a result of the CRVS boundary
being inoperable. This allowance requires proceduralized
compensatory measures to protect the operators during that time
period. Although this change does provide for an increase in the
allowed time for continued plant operation, its acceptability is
based on the low probability of any design basis accident during
that time period and the protection provided by the compensatory
measures that would be established. The design and operation of the
control room ventilation system is not affected by this change.
The proposed changes also provide for limiting the applicability
of SR 3.7.12.3, which verifies the operability of the ABVS on an SI
signal. The limitation is imposed only when the ABVS is aligned and
operating in its safety function configuration. Since the ABVS is
already performing its safety function when it is in this condition,
verifying the automatic capability to transfer to this configuration
is unnecessary. Since this limitation is only during periods where
the ABVS is in its safety function configuration, it does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
The proposed changes also provide for limiting the required
operability of the FHBVS based on a minimum time period that all
fuel assemblies in the fuel pool have not been part of a critical
core. This change does reduce the current operability requirements
for the FHBVS by limiting these requirements to the period when the
system would be required to mitigate the radiological consequences
of an accident to acceptable limits. However, the design and
operation of the FHBVS is not affected by this change.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The charcoal adsorber sample laboratory testing protocol
accurately demonstrates the required performance of the adsorbers in
the CRVS and ABVS following a design basis accident or in the FHBVS
following a fuel handling accident outside containment. The changes
in charcoal testing acceptance criteria and frequency will not
affect system performance or operation. They will continue to ensure
that the charcoal will perform its safety function. The
decontamination efficiencies used in the offsite and control room
dose analyses are not affected by this change. Therefore the offsite
and control room dose analyses are not affected by this change, and
offsite and control room doses will remain within the limits of 10
CFR 100 and 10 CFR 50, Appendix A, GDC [General Design Criterion]
19. Although there is a reduction in the safety factor provided by
the previous testing protocol, the revised testing protocol follows
current industry standards. These standards ensure adequate margin
exists and that the charcoal will perform its design basis function.
As a result, there is no significant reduction is [in] a margin of
safety.
The proposed changes also provide for two trains of the CRVS to
be inoperable for up to
[[Page 37581]]
24 hours as a result of the control room boundary being inoperable.
Although this change does provide for an increase in the allowed
time for continued plant operation under certain conditions, its
acceptability is based on a low probability of any design basis
accident occurring during that time period and the added protection
provided by the compensatory measures that would be established. The
increase in inoperability could be considered to be a decrease in
the margin of safety of this system. However, based on the low
probability of a concurrent accident requiring system operability
during the completion time for this condition and the ability of the
compensatory measures to restore the boundary before it is needed if
an accident occurs, this potential reduction in safety margin is not
considered to be significant.
The proposed changes also provide for limiting the applicability
of SR 3.7.12.3, which verifies the operability of the ABVS on a SI
signal. The limitation is imposed only when the ABVS is aligned and
operating in its safety function configuration. Since the ABVS is
already performing its safety function when it is in this condition,
verifying the automatic capability to transfer to this configuration
is unnecessary. Since this limitation is only during periods where
the ABVS is already in its safety function configuration, the margin
of safety is actually increased because the ABVS does not have to
change configuration as a result of an accident to perform its
safety function.
The proposed changes also provide for limiting the required
operability of the FHBVS based on a minimum time period (``recently
irradiated fuel'') that all fuel assemblies in the fuel pool have
not been part of a critical core. This change does reduce the
current operability requirements for the FHBVS by limiting
operability to the period when the system would be required to
mitigate the radiological consequences of an accident to acceptable
limits. This proposed change creates the potential for increased
dose in the control room and at the site boundary due to a FHA
outside containment. However, the new analysis demonstrates that the
resultant doses are well within the Regulatory Guide (RG) 1.183
limits and within the GDC 19 limits. In the case of the offsite dose
values, they remain within the RG 1.183 limits, which is considered
acceptable. Based on this, the margin of safety is not significantly
reduced.
In the new FHA analysis, the offsite and control room doses due
to a FHA outside containment have been evaluated using conservative
assumptions, such as no credit being taken for the functionality of
either FHBVS train's activated charcoal adsorber sections, the
control room ventilation system remains in normal mode with no
charcoal filtration available, and all airborne activity caused by
the FHA is released at a linear rate over two hours. These
conservative assumptions ensure the results of the calculation
bounds the expected dose. The normal availability of the fuel
handling building and control room filtration systems will reduce
the potential control room and offsite doses in the event of a FHA,
and provides additional margin to the calculated doses.
Therefore, the proposed changes do not involve a significant
reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, PO Box 7442, San Francisco, CA 94120.
NRC Section Chief: Stephen Dembek.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: May 29, 2003.
Description of amendment requests: The proposed changes to the
technical specifications would extend the completion time for restoring
an inoperable diesel generator from 7 days to 14 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes revise the Technical Specification (TS)
3.8.1 completion times for Required Actions A.2 and B.4 associated
with the diesel generators (DGs). The proposed changes allow an
extension of the current TS completion time from 7 days to 14 days
for an inoperable DG.
The proposed changes do not affect the design of the DGs, the
operational characteristics or function of the DGs, the interfaces
between the DGs and other plant systems, or the reliability of the
DGs. Required Actions and the associated completion times are not
initiating conditions for any accident previously evaluated, and the
DGs are not initiators of any previously evaluated accidents. The
DGs mitigate the consequences of previously evaluated accidents
including loss of offsite power. The consequences of a previously
analyzed event will not be significantly affected by the extended DG
completion time since the DGs will continue to be capable of
performing their accident mitigation function as assumed in the
accident analysis. Thus the consequences of accidents previously
analyzed are unchanged between the existing TS requirements and the
proposed changes. The consequences of an accident are independent of
the time the DGs are out of service as long as adequate DG
availability is assured. The proposed changes will not result in a
significant decrease in DG availability so that the assumptions
regarding DG availability are not impacted.
To fully evaluate the effect of the proposed DG completion time
extension, probabilistic risk assessment methods and a deterministic
analysis were utilized. The results of the analysis show no
significant increase in core damage frequency and large early
release frequency.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different accident from any accident previously evaluated.
The proposed changes do not involve a change in the design,
configuration, or method of operation of the plant. The proposed
changes will not alter the manner in which equipment operation is
initiated, nor will the function demands on credited equipment be
changed. No alteration in the procedures which ensure that the plant
remains within analyzed limits is being proposed, and no change is
being made to the procedures relied upon to respond to an off-normal
event. As such, no new failure modes are being introduced.
Therefore, the proposed changes do not create the possibility of
a new or different accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed 14 day DG completion time is based upon both a
deterministic evaluation and a risk-informed assessment. The
availability of offsite power coupled with the availability of the
other DGs in the affected unit, the unit auxiliary feedwater pumps,
and all auxiliary saltwater trains (including the cross-tie) and
utilization of the Online Risk Management Program while a DG is
inoperable, provide adequate compensation for the potential small
incremental increase in plant risk of the extended DG completion
time. In addition, the increased availability of the DGs during
refueling outages provides a reduction in plant risk during shutdown
periods.
The risk assessment performed to support this license amendment
request concluded that the increase in plant risk is small and
consistent with the NRC's Safety Goal Policy Statement, ``Use of
Probabilistic Risk Assessment Methods in Nuclear Activities: Final
Policy Statement,'' Federal Register, Volume 60, p. 42622, August
16, 1995 and guidance contained in [* * *] Regulatory Guides (RG)
1.174, ``An Approach for Using Probabilistic Risk Assessment in
Risk-Informed Decisions on Plant-Specific Changes to the Licensing
Basis,'' dated July 1998 and RG 1.177, ``An Approach for Plant-
Specific, Risk-Informed Decision making: Technical Specifications,''
dated August 1998. Together, the deterministic evaluation and the
risk-informed assessment provide high assurance of the capability to
provide power to the engineered safety feature buses during the
proposed 14 day DG completion time.
[[Page 37582]]
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: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: May 6, 2003.
Description of amendment request: The proposed amendments would
delete SSES 1 and 2 Technical Specifications (TSs) 3.3.1.3,
``Oscillation Power Range Monitor (OPRM) Instrumentation,'' and revise
TS 3.4.1, ``Recirculation Loops Operating.'' These changes would
reverse approved TS Amendment Nos. 184 (Unit 1) and 158 (Unit 2) dated
July 30, 1999, that are not yet implemented, which effectively results
in no change to the current SSES 1 and 2 operation. Extension of the
implementation date was needed to provide time to address continuing
hardware and software deficiencies with the OPRM system. The extension
of the implementation date until November 1, 2001, was approved by
Amendment Nos. 187 (Unit 1) and 161 (Unit 2) dated June 2, 2000. A
second extension of the implementation date until November 1, 2003, was
approved by Amendment Nos. 196 (Unit 1) and 172 (Unit 2) dated October
29, 2001. This deferral was based on a Title 10 Code of Federal
Regulations (10 CFR), part 21, report issued by General Electric
Company on August 31, 2001, which identified a non-conservative
deficiency in the OPRM trip setpoint methodology. The licensee stated
that the OPRM system cannot be declared OPERABLE until a revised NRC-
approved methodology providing a valid basis for the trip setpoints is
available and adopted for the SSES 1 and 2 OPRM systems. The
implementation requirements associated with Amendment Nos. 187, 161,
196 and 172 would also be superceded with this proposed amendment. The
proposed amendment would formally reinstate the requirements currently
governing operation, which define appropriately conservative
restrictions to plant operation and operator response to thermal
hydraulic instability events.
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?
Response: No.
The OPRM system is not an initiator to any accident sequence
analyzed in the Final Safety Analysis Report (FSAR). The changes do
not involve a physical change to structures, systems, or components
(SSCs) since the RPS [reactor protection system] trip function has
not been installed and does not alter the method of operation or
control of SSCs since the OPRM system has not been declared
OPERABLE. The current assumptions in the safety analysis regarding
accident initiators and mitigation of accidents (including assumed
protection of fuel design limits) are unaffected by these changes.
No additional failure modes or mechanisms are being introduced and
the likelihood of previously analyzed failures remains unchanged.
Operation in accordance with the proposed Technical
Specification (TS) ensures that the protection from thermal
hydraulic instabilities remains as previously evaluated and the
protection for fuel design limits remain as described in the FSAR.
Therefore, the mitigative functions will continue to provide the
protection assumed by the existing analysis.
Therefore, this 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 does not involve a physical alteration of
the plant. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. There
are no setpoints affected by this change at which protective or
mitigative actions are initiated. This change will not alter the
manner in which equipment operation is initiated, nor will the
functional demands on credited equipment be changed. No alterations
in the procedures that ensure the plant remains within analyzed
limits are being proposed, and no changes are being made to the
procedures relied upon to respond to an off-normal event as
described in the FSAR. As such, no new failure modes are being
introduced. The change does not alter assumptions made in the safety
analysis and licensing basis.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The proposed change is acceptable because the
required protection from thermal hydraulic instabilities remains as
previously evaluated and the protection for fuel design limits
remain as described in the FSAR. Operation in accordance with the
proposed TS ensures that the margin of safety is maintained.
Therefore, the change does not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: Richard J. Laufer.
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 14, 2003.
Description of amendment request: The proposed amendment revises
surveillance requirement 4.6.2.1 for demonstrating operability of
containment spray system spray nozzles.
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 license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The Containment Spray System is not considered an initiator of
any analyzed event. The proposed change does not have a detrimental
impact on the integrity of any plant structure, system, or component
that may initiate an analyzed event. The proposed change will not
alter the operation or otherwise increase the failure probability of
any plant equipment that can initiate an analyzed accident.
This change does not affect the plant design. There is no
increase in the likelihood of formation of significant corrosion
products. Due to their location at the top of the containment,
introduction of foreign material into the spray headers is unlikely.
Foreign material introduced during maintenance activities would be
the most likely source for obstruction, and verification following
such maintenance would confirm the nozzles remain unobstructed.
Consequently, there is no significant increase in the
probability of an accident previously evaluated.
The Containment Spray System is designed to address the
consequences of a LOCA [loss
[[Page 37583]]
of coolant accident]. The Containment Spray System is capable of
performing its function effectively with the single failure of any
active component in the system, any of its subsystems, or any of its
support systems. A plugged nozzle would have negligible impact on
the capability of the Containment Spray System to respond to a Loss
of Coolant Accident.
Therefore, the consequences of an accident previously evaluated
are not significantly affected by the proposed change.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated? The proposed change will not physically alter the plant
(no new or different type of equipment will be installed) or change
the methods governing normal plant operation. Therefore, this change
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
The system is not susceptible to corrosion-induced obstruction
or obstruction from sources external to the system. Maintenance
activities that could introduce foreign material into the system
would require subsequent verification to ensure there is no nozzle
blockage. The spray header nozzles are expected to remain unblocked
and available in the event that the safety function is required.
Therefore, the capacity of the system would remain unaffected.
Hence, this change does not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: Robert A. Gramm.
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 22, 2003.
Description of amendment request: The proposed amendments revise
Technical Specification (TS) 3.4.2.2, ``Reactor Coolant System,'' to
relax the lift setting tolerance of the pressurizer safety valves from
+/-2 percent to +2 percent, -3 percent.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed TS change takes credit for the assumptions made in
the reanalysis of the rod withdrawal from power event already
evaluated in the UFSAR [Updated Final Safety Analysis Report].
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed TS change takes credit for the assumptions made in
the reanalysis of the rod withdrawal from [the] power event already
evaluated in the UFSAR. Therefore, the change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel and fuel cladding, reactor
coolant pressure boundary, and containment structure) to limit the
level of radiation dose to the public. The proposed TS change takes
credit for the assumptions made in the reanalysis of the rod
withdrawal from power event already evaluated in the UFSAR. That
analysis demonstrated that the fuel design limits were maintained by
the reactor protection system since the DNBR [Departure from
Nucleate Boiling Ratio] was maintained above the limit value.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
Based on the above, STPNOC concludes that the proposed amendment
involves no significant hazards consideration under the standards
set forth in 10 CFR 50.92 and, accordingly, a finding of ``no
significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: Robert A. Gramm.
Tennessee Valley Authority (TVA), Docket No. 50-328, Sequoyah Nuclear
Plant (SQN), Unit 2, Hamilton County, Tennessee
Date of amendment request: June 5, 2003 (TSC 03-08).
Description of amendment request: The proposed amendment would
revise the reactor coolant system (RCS) heatup and cooldown curves
(pressure-temperature (P-T) limits). The revision replaces the P-T
limits that are currently analyzed for 14.5 Effective Full Power Years
(EFPYs) with new limits analyzed for 32 EFPYs. In addition, the
amendment includes corresponding changes to the Technical Specification
(TS) figure associated with the Low Temperature Over Pressure
Protection and the TS Bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed revision does not affect plant equipment, test
methods or operating practices. The modification to SQN TSs is
consistent 10 CFR 50, Appendix G in conjunction with alternative
methods provided in American Society of Mechanical Engineers (ASME)
Code Case N-640, ``Alternative Requirement Fracture Toughness for
Development of P-T Limit Curves for ASME Section XI, Division 1.''
The proposed change continues to provide controls for safe operation
within the required limits. The proposed changes do not contribute
to events or assumptions associated with postulated design basis
accidents (DBA). The proposed revisions continue to maintain the
required safety functions. Accordingly, the probability of an
accident or the consequences of an accident previously evaluated is
not increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed revision is not the result of changes to plant
equipment, test methods, or operating practices. The proposed
revision to the SQN Unit 2 P-T limits continues to ensure that
conservative fracture toughness margins are maintained to protect
against reactor pressure vessel failure. In addition, SQN's current
setpoints for low-temperature overpressure protection were evaluated
and are bounding for the proposed 32 EFPY P-T limits. The updated P-
T limits are based on NRC approved methodology in conjunction with
alternative methods provided in American Society of Mechanical
Engineers (ASME) Code Case N-640, ``Alternative Requirement Fracture
Toughness for Development of P-T Limit Curves for ASME Section XI,
Division 1.''
The reactor vessel P-T limits are operational limits and are not
considered to be contributors to the generation of postulated
accidents. The safety functions of the associated systems remain
unchanged and do not affect the assumptions of DBAs. The operational
limits continue to be governed within the TSs. Accordingly, the
proposed change does not create the possibility of a new or
different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
[[Page 37584]]
No. TVA's proposed TS amendment provides revised reactor
pressure vessel P-T limits that are within the design capabilities
of the pressure control systems for protection of the RCS. The
limits are based on conservative design margins that ensure that
plant operation is within the design capacity of the reactor vessel
materials. Accordingly, the function of the RCS to provide a fission
product barrier is not compromised.
TVA's proposed change to revise P-T limits does not result in a
change to system design features. The proposed change does not
affect plant conditions that result in precursors to accidents or
cause degradation of accident mitigation systems. The plant system
safety functions are not altered by the proposed change.
The proposed changes allow plant operation with different P-T
limits while continuing to retain conservative margins for assuring
integrity of the reactor vessel and the RCS. Consequently, the
proposed TS revisions do not significantly reduce 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, TN 37902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: June 5, 2003 (TSC 03-09).
Description of amendment request: The proposed change to the
Updated Final Safety Analysis Report (UFSAR) would amend the design and
licensing basis to identify that operator action may be necessary to
ensure containment design pressure is not exceeded subsequent to a high
energy line break (HELB) such as loss-of-coolant-accident.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated.
No. The procedure changes/additions being implemented to
mitigate a SCSA [station control and service air] leak in
containment will only be used following a HELB in containment, a
consequential rupture of an SCSA line and a failure of the outboard
CIV [containment isolation valve] on the SCSA containment supply.
Operators isolate the SCSA leak on the accident unit by either
manually closing a valve upstream of the stuck-open CIV or by
shutting down the station air compressors. If the station air
compressors are shut down prior to performing an emergency shut down
of the non-accident unit or if an operator error results in an
isolation of the control air supply to the non-accident unit, then
at-worst, a UFSAR Condition II event is induced on the non-accident
unit. For example a reactor trip from full power or a loss of normal
feedwater--loss of control air to the feedwater regulator valves
resulting in a loss of normal feedwater to the non-accident unit.
A UFSAR Condition II event has a frequency of one per year.
Therefore, the proposed procedure changes/additions, including the
potential for operator error do not result in more than a minimal
increase in a previously evaluated Condition II event (1+1/40 =
1.025 less than 10 percent increase).
The operator actions being implemented to mitigate a SCSA leak
in containment are performed after the occurrence of an accident on
primarily non-safety-related systems, structures or components
[SSCs] so they do not increase the likelihood of the occurrence of a
malfunction of equipment previously evaluated in the UFSAR.
The air operated containment isolation valve is assumed to fail
open due to single failure criteria and, containment isolation/
integrity is maintained by the inboard check valve. The containment
boundary is unaffected by the operator actions being implemented to
mitigate a SCSA leak in containment. Therefore, the consequences of
all accidents previously evaluated in the UFSAR remain unchanged.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated.
No. This change implements new manual actions for accident
failure modes not previously evaluated in the UFSAR. The manual
actions are required to ensure containment design pressure is not
exceeded.
Operators isolate the SCSA leak on the accident unit by either
manually closing an upstream isolation valve on the accident unit's
SCSA containment supply or by shutting down the station air
compressors. The operator actions being implemented have been
determined to meet the criteria for safety-related operator actions
in NRC Information Notice (IN) 97-78/ANS-58.8 and; therefore, there
are no credible operator actions which would prevent isolation of a
SCSA leak prior to containment design pressure being exceeded. If
the station air compressors are shut down prior to performing an
emergency shutdown of the non-accident unit or if an operator error
results in an isolation of the SCSA supply to the non-accident unit,
then at-worst, a UFSAR Condition II event occurs. Because UFSAR
Condition II events have been previously identified, the operator
actions being added under this change do not create the possibility
of an accident of a different type than previously evaluated.
The operator actions being implemented to mitigate a SCSA leak
in containment are performed after the occurrence of an accident on
primarily non-safety-related SSCs so they do not create a
possibility for a malfunction of an SSC important to safety with a
different result than previously evaluated in the UFSAR.
3. Does the proposed change involve a significant reduction in a
margin of safety.
No. The established limits for the fuel, reactor vessel or
containment are not affected by the addition of operator actions to
isolate a SCSA leak inside containment. Isolation of the air leak
within two hours of a large break loss-of-coolant accident (LBLOCA)
prevents containment pressure exceeding the peak calculated
pressure. Consequently, this change does not represent a 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, TN 37902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority (TVA), Docket No. 50-390, Watts Bar Nuclear
(WBN) Plant, Unit 1, Rhea County, Tennessee
Date of amendment request: May 14, 2003.
Description of amendment request: The proposed amendment would
allow an alternate Westinghouse methodology for the measurement of
reactor coolant system (RCS) total flow rate via measurement of the RCS
elbow tap differential pressures. TVA stated that this methodology is
similar to that reviewed and approved by the NRC for other utilities.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. TVA's evaluation for WBN Unit 1 determined that the
probability of an accident will not increase since adequate RCS flow
will still be assured. Sufficient margin exists to account for all
reasonable instrument uncertainties; therefore, no changes to
installed equipment or hardware in the plant are required, thus the
probability of an accident occurring remains unchanged. The initial
conditions for all accident
[[Page 37585]]
scenarios modeled are the same and the conditions at the time of
trip, as modeled in the various safety analyses are the same.
Therefore, the consequences of an accident will be the same as those
previously analyzed.
Therefore, since the actual plant configuration, performance of
systems, and initiating event mechanisms are not being changed, TVA
has concluded that 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?
No. There are no changes in operation of the plant that could
introduce a new failure mode. No new accident scenarios have been
identified. Operation of the plant will be consistent with that
previously modeled, i.e., the time of reactor trip in the various
safety analyses is the same, thus plant response will be the same
and will not introduce any different accident scenarios that have
not been evaluated.
Therefore, TVA concludes that this 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?
No. The proposed change reflects changes due to the method used
to verify RCS flow at the beginning of each cycle. However, no
changes to the Safety Analysis assumptions were required; therefore,
the margin of safety will remain the same. Therefore, TVA concludes
that the proposed change does not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, (301) 415-4737 or by email to [email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: December 16, 2002, as
supplemented on April 1, 2003.
Brief description of amendment: The amendment revised the Technical
Specifications, changing the safety limit minimum critical power ratio
(SLMCPR) from 1.11 to 1.09 for both four- or five-recirculation-loop
operation, and from 1.12 to 1.10 for three-recirculation-loop
operation. It also added a paragraph to explain that the lower SLMCPR
values are due primarily to an improved treatment of the power
distribution uncertainty.
Date of issuance: June 5, 2003.
Effective date: June 5, 2003 and shall be implemented within 30
days of issuance.
Amendment No.: 238.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 21, 2003 (68 FR
2799).
The April 1, 2003, letter provided clarifying information within
the scope of the original application and did not change the staff's
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated June 5, 2003.
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: December 2, 2002, as
supplemented by letter dated April 14, 2003.
Brief description of amendments: The amendments revise the
Technical Specifications for Administrative Controls in Section 5.0
concerning Responsibility, Unit Staff, Unit Staff Qualifications, and
controls for the High Radiation Area.
Date of issuance: June 6, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 213 and 194.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 7, 2003 (68 FR
800).
The supplement dated April 14, 2003, provided clarifying
information that did not change the scope of the December 2, 2002,
application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 6, 2003.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: December 30, 2002, and its
supplement dated April 28, 2003.
Brief description of amendment: The amendment revises Technical
Specification (TS) 2.1.1.2, ``Minimum Critical Power Ratio Safety Limit
[[Page 37586]]
(MCPRSL)'' to support operating during Cycle 17. Cycle 17 is the first
cycle of operation with a mixed core of ABB/CE/Westinghouse SVEA-96
fuel and Framatome ANP AtriumTM-10 reload fuel. The
amendment also revises Surveillance Requirement (SR) 3.3.1.3.2--the low
power range monitor (LPRM) calibration frequency specified in the TS
for the oscillation power range monitor. This change corrects an
inconsistency between the LPRM calibration frequency specified in SR
3.3.1.3.2 and SR 3.3.1.1.7, ``Reactor Protection System (RPS)
Instrumentation.''
Date of issuance: June 2, 2003.
Effective date: June 2, 2003, and shall be implemented before the
plant restarts after completion of Refueling Outage 16.
Amendment No.: 186.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 18, 2003 (68
FR 7815).
The April 28, 2003, supplemental letter provided additional
clarifying information, did not change the scope of the application as
originally noticed, and did not change the staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 2, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of application for amendment: October 4, 2002.
Brief description of amendment: Change the Technical Specifications
(TSs) by extending the primary containment integrated leak rate testing
interval from 10 years to no longer than approximately 10.6 years, on a
one-time basis.
Date of issuance: June 2, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 215.
Facility Operating License No. DPR-28: Amendment revised the TSs.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68736).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated June 2, 2003.
No significant hazards consideration comments received: No.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: December 23, 2002, as supplemented
January 24 and April 21, 2003.
Description of amendment request: The amendment relocates, intact,
Technical Specification (TS) 6.2.3, ``Independent Technical Reviews;''
TS 6.4, ``Review and Audit;'' TS 6.7.2 through 6.7.5 (specific
descriptions of the procedure review and approval process); and TS 6.9,
``Records Retention'' to the Operational Quality Assurance Program. The
amendment also changes the title of the senior onsite official from
``Executive Vice President and Chief Nuclear Officer'' to ``Site Vice
President,'' revises the 10 CFR 20 references in the TSs to bring them
into consistency with 10 CFR 20, and makes other minor editorial
changes.
Date of issuance: June 6, 2003.
Effective date: As of its date of issuance, and shall be
implemented within 60 days.
Amendment No.: 88.
Facility Operating License No. NPF-86: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 18, 2003 (68
FR 7817).
The April 21, 2003, letter provided clarifying information that did
not change the initial proposed no significant hazards consideration
determination or expanded the application beyond the scope of the
original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 6, 2003.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: July 10, 2002, as supplemented by letter
dated April 16, 2003.
Brief description of amendment: The amendment replaces the fire
protection requirements contained in Facility Operating License (FOL)
Section 2.C.(4) with the standard fire protection FOL condition
recommended by Generic Letter 86-10, Section F, adapted to Cooper
Nuclear Station.
Date of issuance: June 5, 2003.
Effective date: June 5, 2003.
Amendment No.: 199.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 7, 2003 (68 FR
808).
The supplement dated April 16, 2003, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 5, 2003.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station, Unit No. 1, Oswego County, New York
Date of application for amendment: October 7, 2002, as supplemented
by letter dated March 24, 2003.
Brief description of amendment: The amendment (1) adds a new
Surveillance Requirement (SR) 4.0.3 to extend the delay period, up to
24 hours or up to the limit of the specified frequency, whichever is
greater, before entering a Limiting Condition for Operation following a
missed surveillance; (2) adds a new SR 4.0.1 to define general
conditions for use of SRs; and (3) makes various editorial and
administrative changes.
Date of issuance: June 3, 2003.
Effective date: June 3, 2003, to be implemented within 60 days of
issuance.
Amendment No.: 182.
Facility Operating License No. DPR-63: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68739) and April 29, 2003 (68 FR 22748).
The supplement expanded the scope of the application, and was
addressed by the second notice. The Commission's related evaluation of
the amendment is contained in a Safety Evaluation dated June 3, 2003.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: June 11, 2002.
Brief description of amendments: These amendments revise Technical
Specification 3.1.8, ``Physics Tests Exceptions--Mode 2,'' to correct
an error in the numbering of a function. Specifically, the reference in
Limiting Condition for Operation 3.1.8 to Function 17.d has been
changed to Function 17.e.
[[Page 37587]]
Date of issuance: June 3, 2003.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 208 & 213.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 3, 2002 (67
FR 56325).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 3, 2003.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: February 6, 2003.
Brief description of amendments: The amendments revise Surveillance
Requirements (SRs) 3.3.1.2 and 3.3.1.3 of the Technical Specifications
on the reactor trip system instrumentation. The proposed changes to SR
3.3.1.2 move Note 1 to the body of the SR, replace the reference to
nuclear instrumentation system channel output by a reference to power
range channel output, and delete the reference to the absolute
difference. The change to SR 3.3.1.3 is editorial.
Date of issuance: June 2, 2003.
Effective date: June 2, 2003, and shall be implemented within 60
days of the date of issuance.
Amendment Nos.: Unit 1-157; Unit 2-157.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 15, 2003 (68 FR
18282).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 2, 2003.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket No. 50-323, Diablo Canyon
Nuclear Power Plant, Unit No. 2, San Luis Obispo County, California
Date of application for amendments: March 3, 2003, and its
supplement dated March 5, 2003.
Brief description of amendments: The amendment authorizes revisions
to the Final Safety Analysis Report (FSAR) Update to incorporate the
NRC approval of a probability of detection of 1.0 to one bobbin
indication, contained in Diablo Canyon Nuclear Power Plant (DCPP) Unit
2 steam generator 4 tube at row 44, column 45 at the second tube
support plate on the hot leg side, for the beginning of cycle voltage
distribution for the DCPP Unit 2 Cycle 12 operational assessment. In a
Federal Register notice dated April 15, 2003 (68 FR 18284), the NRC
described the amendment request as follows:
The proposed license amendment would revise Technical
Specification (TS) 5.5.9, ``Steam Generator Tube Surveillance
Program,'' and TS 5.6.10, ``Steam Generator Tube Inspection
Report,'' for Diablo Canyon Power Plant (DCPP) Unit 2, to apply a
probability of detection (POD) of 1.0 to the bobbin indication in
the steam generator (SG) 4 tube at row 44, column 45 at the second
tube support plate (TSP) on the hot leg side (R44C45-2H) for the
beginning of cycle (BOC) voltage distribution for the DCPP Unit 2
BOC Cycle 12 operational assessment.
The change from a TS to an FSAR revision resulted from the March 5,
2003, supplement and is not substantial in that the technical issues
and no significant hazards consideration determination remain the same.
Date of issuance: June 3, 2003.
Effective date: June 3, 2003, and shall be implemented within 30
days of the date of issuance. The implementation of the amendment
includes the incorporation into the FSAR Update the changes discussed
above, as described in the licensee's application dated March 3, 2003,
its supplement dated March 5, 2003, and evaluated in the staff's safety
evaluation attached to the amendment.
Amendment No.: 158.
Facility Operating License No. DPR-82: The amendment authorized
revision of the FSAR Update.
Date of initial notice in Federal Register: April 15, 2003 (68 FR
18284).
The March 5, 2003, supplemental letter provided additional
clarifying information, did not expand the scope of the application as
originally noticed, and did not change the NRC staff's original
proposed no significant hazards consideration determination. The
Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated June 3, 2003.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendments: October 31, 2002.
Brief description of amendments: These amendments revised the
Technical Specifications, Section 3.7.6, ``Main Turbine Bypass
System,'' to change the requirement for operability of the main turbine
bypass system bypass valves. Specifically, Surveillance Requirement
3.7.6 would be revised to test only each required turbine bypass valve
every 31 days.
Date of issuance: May 29, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 210 and 185.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 24, 2002 (67
FR 78524).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 29, 2003.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendments: October 30, 2002.
Brief description of amendments: These amendments revise Technical
Specifications Section 5.5.7, ``Ventilation Filter Testing Program,''
to change the control room emergency outside air supply system
(CREOASS) maximum allowed filter train pressure drop from <9.1 inches
water gage (wg) to <7.3 inches wg.
Date of issuance: May 29, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 211 and 186.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 24, 2002 (67
FR 78523).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 29, 2003.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendments: March 3, 2003.
Brief description of amendments: The amendments delete Technical
Specification (TS) 5.5.3, ``Post Accident Sampling,'' and thereby
eliminate the requirements to have and maintain the post-accident
sampling systems. The amendments also address related changes to TS
5.5.2, ``Primary Coolant Sources Outside Containment.''
Date of issuance: June 3, 2003.
[[Page 37588]]
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 212 and 187.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 29, 2003. (68 FR
22752).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 3, 2003.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendments: October 31, 2002.
Brief description of amendments: These amendments revised Technical
Specifications, Section 3.3.6.1, ``Primary Containment Isolation
Instrumentation,'' to add an ACTIONS Note allowing intermittent
opening, under administrative control, of penetration flow paths that
are isolated. Additionally, these amendments revised TSs Section
3.3.6.1 to breakout the traversing incore probe system isolation as a
separate isolation Function with an associated Required Action to
isolate the penetration within 24 hours rather than immediately
initiating a unit shutdown.
Date of issuance: June 5, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 213 and 188.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 24, 2002 (67
FR 78523).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 5, 2003.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of application for amendments: May 8, 2002, as supplemented by
letters dated November 26, 2002, and April 10, 2003.
Brief description of amendments: The amendments revise the Reactor
Core Safety Limits curve in Technical Specifications (TS) Figure 2.1.1-
1, and the Over Temperature Delta Temperature (OTDT) and Over Power
Delta Temperature (OPDT) reactor trip functions described in TS Table
3.3.1-1. These changes will provide Vogtle Electric Generating Plant
(VEGP), Units 1 and 2 with increased operating margins that will
increase the OTDT and OPDT setpoints to account for hot leg temperature
fluctuations that are part of the VEGP Setpoint Margin Recovery
Program.
Date of issuance: June 4, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 128/106.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 23, 2002.
The supplements dated November 26, 2002, and April 10, 2003,
provided clarifying information that did not change the scope of the
May 8, 2002, application nor the initial proposed no significant
hazards consideration determination. The Commission's related
evaluation of the amendments is contained in a Safety Evaluation dated
June 4, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant,
Units 1and 2, Hamilton County, Tennessee
Date of application for amendment: October 4, 2002, as supplemented
February 19, 2003, and May 19, 2003.
Brief description of amendment: The amendments revise Technical
Specification (TS) 6.8.4.h, Containment Leakage Rate Testing Program,
to allow the licensee to postpone its Appendix J, Type A, Containment
Integrated Leak Rate Test (ILRT) for 5 years. Specifically, for Unit 1
the performance of the spring 2003 ILRT may be deferred up to 5 years
but no later than spring 2008, and for Unit 2 performance of the fall
2003 ILRT may be deferred up to an additional 3.5 years but no later
than spring 2007. In Amendment No. 265 to the Facility Operating
License No. DPR-79 for SQN, Unit 2, TS 6.8.4.h was revised to allow the
licensee to postpone the ILRT one cycle (i.e., 1.5 years) from spring
2002. Therefore, the total deferral for SQN, Unit 2 from the original
requirement to perform a ILRT in spring 2002 will be up to 5 years.
Date of issuance: May 29, 2003.
Effective date: As of the date of issuance and shall be implemented
within 45 days of issuance.
Amendment Nos.: 287 and 276.
Facility Operating License No. DPR-77: Amendment revises the
technical specifications.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5681).
The February 19, and May 19, 2003, letters provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the scope of the
application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 29, 2003.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: November 19, 2002, as supplemented by
letters dated February 5 and May 5, 2003.
Brief description of amendments: The amendments revise Appendix B
to the Facility Operating License, Environmental Protection Plan (EPP),
to replace references to the U.S. Environmental Protection Agency's
National Pollutant Discharge Elimination System expired permit. The
amendments also contain minor changes to the EPP to be consistent with
the provisions of the current Texas Pollutant Discharge Elimination
System permit and the Final Environmental Statement--Operating License
Stage, and consolidate the Unit 1 and Unit 2 EPPs into a single
document.
Date of issuance: May 29, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 104 and 104.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Facility Operating License, Appendix B, ``Environmental
Protection Plan.''
Date of initial notice in Federal Register: December 24, 2002 (67
FR 78524).
The supplemental letters provided clarifying information that did
not change the scope of the original Federal Register notice or the
original no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 29, 2003.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: March 21, 2003.
[[Page 37589]]
Brief description of amendment: The amendment revises paragraphs in
Section 5.0, ``Administrative Controls,'' of the Technical
Specifications to allow the use of generic personnel titles in place of
plant-specific personnel titles and requires either the operations
manager or assistant operations manager to hold a senior reactor
operator license.
Date of issuance: June 3, 2003.
Effective date: June 3, 2003, and shall be implemented within 30
days of the date of issuance, including the incorporation of the Final
Safety Analysis Report changes described in the licensee's application
dated March 21, 2003, and the staff's Safety Evaluation for this
amendment.
Amendment No.: 155.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 16, 2003 (68 FR
18714).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 3, 2003.
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 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, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland. Publicly available
records will be accessible from the Agencywide Documents Assess and
Management System's (ADAMS) Public Electronic Reading Room on the
Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html.
If you do not have access to ADAMS or if there are problems in
accessing the documents located in ADAMS, contact the NRC Public
Document Room (PDR) Reference staff at 1-800-397-4209, (301) 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 July 24, 2003, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.714, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, and
electronically on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are problems in accessing the
document, contact the PDR Reference staff at 1-800-397-4209, (301) 415-
4737, or by e-mail to [email protected]. 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
[[Page 37590]]
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 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 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of the continuing disruptions in delivery of mail to United
States Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to (301) 415-
1101 or by e-mail to [email protected]. A copy of the petition for
leave to intervene and request for hearing should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and because of continuing disruptions in
delivery of mail to United States Government offices, it is requested
that copies be transmitted either by means of facsimile transmission to
301-415-3725 or by e-mail to [email protected]. A copy of the
request for hearing and petition for leave to intervene should also be
sent to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: May 1, 2003, as supplemented
May 2 and May 15, 2003.
Brief description of amendments: The amendments modify technical
specification surveillance requirements to provide an alternative means
of testing the Unit 1 main steam Electromatic relief valves, including
those that provide the automatic depressurization and the low set
relief functions, and provide an alternative means for testing the
Units 1 and 2 dual function Target Rock safety/relief valves.
Date of issuance: May 28, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 216/210.
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. 68 FR 25645, dated May 13, 2003. The notice
provided an opportunity to submit comments on the Commission's proposed
NSHC determination. No comments have been received. The supplements
dated May 2 and May 15, 2003, provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original proposed
NSHC determination. The Commission's related evaluation of the
amendment, finding of exigent circumstances, state consultation, and
final NSHC determination are contained in a Safety Evaluation dated May
28, 2003.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
Dated at Rockville, Maryland, this 16th day of June 2003.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 03-15597 Filed 6-23-03; 8:45 am]
BILLING CODE 7590-01-P