[Federal Register Volume 68, Number 32 (Tuesday, February 18, 2003)]
[Notices]
[Pages 7810-7827]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-3689]
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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
[[Page 7811]]
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, January 24, 2003, through February 6, 2003.
The last biweekly notice was published on February 4, 2003 (68 FR
5668).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the 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 March 20, 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,\1\ 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.
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\1\ The most recent version of title 10 of the Code of Federal
Regulations, published January 1, 2002, inadvertently omitted the
last sentence of 10 CFR 2.714(d) and paragraphs (d)(1) and (d)(2)
regarding petitions to intervene and contentions. For the complete,
corrected text of 10 CFR 2.714(d), please see 67 FR 20884; April 29,
2002.
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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
[[Page 7812]]
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 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, 304-415-4737 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: December 13, 2002.
Description of amendments request: The proposed amendments would
revise Technical Specification 3.5.2, Emergency Core Cooling System--
Operating, by removing the Note that modifies the Limiting Condition
for Operation. The proposed change would remove the requirement to have
the charging pumps operable when thermal power is greater than 80% of
rated thermal power (RTP). The proposed change would also remove
Surveillance Requirement 3.5.2.4 for verifying the required charging
pump flow rate.
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 charging pumps were credited in the previous analysis to
mitigate the consequences of a small-break loss-of-coolant accident
(LOCA) above 80% of rated thermal power (RTP). The charging pumps
were not considered to be an initiator of the accident. The new
analysis for the small-break LOCA does not assume the charging pumps
are initiators of the accident. Therefore, removing the requirement
to maintain the charging pumps operable above 80% RTP and removing
Surveillance Requirement 3.5.2.4 from the Technical Specification
does not involve a significant increase in the probability of an
accident previously evaluated.
The consequence of a small-break LOCA is the potential for
inadequate core cooling and decreased negative reactivity such that
the reactor core is not protected after the design basis event. The
previous analysis for the small-break LOCA above 80% RTP assumed
unborated flow from a single charging pump to ensure there was
adequate cooling flow delivered to the Reactor Coolant System. The
revised small-break LOCA analysis was performed such that flow from
the charging pumps was not credited. Since the charging pump flow is
no longer credited in the small-break LOCA analysis, the proposed
changes do not involve a significant increase in the consequences of
a small-break LOCA.
Therefore, the proposed Technical Specification changes do not
involve a significant increase in 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.
This request[ed] change does not involve a change in the
operation of the plant and no new accident initiation mechanism is
created by the proposed changes. Since the charging pump flow is no
longer credited in the small-break LOCA analysis, the requirement to
have the charging pumps operable above 80% RTP and the charging pump
Surveillance Requirement 3.5.2.4 can be removed from the Technical
Specification. 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. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The safety function of the Emergency Core Cooling System is to
provide core cooling and negative reactivity, to ensure that the
reactor core is protected after design basis events. For a small-
break LOCA, the previous analysis credited flow from the charging
pumps above 80% RTP to supply supplemental cooling flow to the
Reactor Coolant System. Credit for flow from a single charging pump
was only taken for the water inventory.
The revised small-break LOCA analysis was performed using the
newest Nuclear Regulatory Commission accepted versions of the
Westinghouse evaluation models for Combustion Engineering designed
pressurized water reactors. The revised small-break LOCA analysis
incorporated several changes to plant parameters used in the
analysis, one of which was the elimination of the need to credit the
charging pump flow above 80% RTP. Since the charging pump flow is no
longer credited in the small-break LOCA analysis, the requirement to
have the charging pumps operable above 80% RTP and charging pump
Surveillance Requirement 3.5.2.4 can be removed from the Technical
Specification.
The proposed change to Technical Specification 3.5.2 does not
modify any other charging pump requirements specified in the
Technical Requirements Manual (e.g., requirements on charging pump
availability for boration and cooldown remain in effect).
Therefore, the safety function of the Emergency Core Cooling
System is maintained and the margin of safety is not significantly
reduced by the proposed changes.
[[Page 7813]]
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.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: November 12, 2002.
Description of amendments request: The proposed amendments would
revise the Technical Specifications, as necessary, to support an
expansion of the core flow operating range (i.e., Maximum Extended Load
Line Limit Analysis Plus (MELLLA+)). As part of the MELLLA+
implementation, Carolina Power & Light Company would implement the
Detect and Suppress Solution-Confirmation Density (DSS-CD) approach to
automatically detect and suppress neutronic/thermal-hydraulic
instabilities.
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:
10 CFR 50.91(a) states ``At the time a licensee requests an
amendment, it must provide to the Commission its analysis about the
issue of no significant hazards consideration using the standards in
Sec. 50.92.'' The following provides this analysis for the MELLLA+
operating range to a minimum core flow rate of 85% of rated with
120% of the original licensed thermal power.
1. Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The expansion of the core operating range discussed herein will
not significantly increase the probability or consequences of an
accident previously evaluated.
The probability (frequency of occurrence) of a DBA [design-basis
accident] occurring is not affected by the operating range
expansion, because the plant continues to comply with the regulatory
and design basis criteria established for plant equipment (ASME
[American Society of Mechanical Engineers] code, IEEE [Institute of
Electrical and Electronics Engineers] standards, NEMA [National
Electrical Manufacturers Association] standards, Regulatory Guides,
etc.). An evaluation of the probabilistic safety assessments
concludes that the calculated core damage frequencies do not
significantly change due to the MELLLA+ operating range expansion.
Scram setpoints (equipment settings that initiate automatic plant
shutdowns) are established such that there is no significant
increase in scram frequency due to the MELLLA+ operating range
expansion. No new challenge to safety related equipment results from
the MELLLA+ operating range expansion. The changes in consequences
of hypothetical accidents, which occur from operation in the MELLLA+
region, are in all cases insignificant. The MELLLA+ accident
evaluations do not exceed any NRC-approved acceptance limits. The
spectrum of hypothetical accidents and abnormal operational
occurrences has been investigated, and will meet the plant's
currently licensed regulatory criteria. In the area of core design,
for example, the fuel operating limits such as Maximum Average
Planar Linear Heat Generation Rate (MAPLHGR) and Safety Limit
Minimum Critical Power Ratio (SLMCPR) are met, and fuel reload
analyses will show plant transients meet the criteria accepted by
the NRC as specified in [GE Nuclear Energy, ``General Electric
Standard Application for Reactor Fuel,'' NEDE-24011-P-A and NEDE-
24011-P-A-US, (latest approved revision)]. Challenges to fuel (ECCS
[emergency core cooling system] performance) are evaluated, and
shown to still meet the criteria of 10 CFR 50.46, Appendix K,
Regulatory Guide 1.70, and UFSAR [Updated Final Safety Analysis
Report] Section 6.3. Challenges to the containment have been
evaluated, and the containment and its associated cooling systems
meet 10 CFR 50 Appendix A Criterion 38, Long Term Cooling, and
Criterion 50, Containment. Radiological release events (accidents)
have been evaluated, and shown to meet the regulatory limits of 10
CFR 50.67.
2. Will the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The MELLLA+ operating range expansion will not create the
possibility of a new or different kind of accident from any accident
previously evaluated. Equipment that could be affected by MELLLA+
has been evaluated and no new operating mode, safety related
equipment lineup, accident scenario, or equipment failure mode was
identified. The full spectrum of accident considerations, defined in
the UFSAR, has been evaluated, and no new or different kind of
accident has been identified. The MELLLA+ operating range expansion
uses fully developed technology, and applies it within the
capabilities of existing plant equipment. The technology includes
NRC approved codes, standards and methods applied in accordance with
existing regulatory criteria.
3. Will the change involve a significant reduction in a margin
of safety?
The MELLLA+ operating range expansion will not involve a
significant reduction in a margin of safety. The calculated loads on
all affected structures, systems and components have been shown to
remain within design allowables for all design basis event
categories. No NRC acceptance criterion is exceeded. The margins of
safety currently included in the design of the plant are not
affected by the MELLLA+ operating range expansion. Because the plant
configuration and response to transients and hypothetical accidents
do not result in exceeding the presently approved NRC acceptance
limits, operation in the MELLLA+ region does not involve a
significant reduction in a margin of safety.
Conclusion: A MELLLA+ operating range expansion to a minimum
core flow rate of 85% of rated with 120% of original licensed
thermal power has been investigated. The BSEP [Brunswick Steam
Electric Plant] licensing requirements have been evaluated and it
has been demonstrated that this MELLLA+ operating range expansion
can be accommodated:
[sbull] Without a significant increase in the probability or
consequences of an accident previously evaluated,
[sbull] Without creating the possibility of a new or different
kind of accident from any accident previously evaluated, and
[sbull] Without exceeding any presently existing regulatory
limits or acceptance criteria applicable to the plant, which might
cause a reduction in a margin of safety.
Having made negative declarations regarding the 10 CFR 50.92
criteria, this assessment concludes that an operating range
expansion to a minimum core flow rate of 85% of rated with 120% of
original licensed thermal power does not involve a Significant
Hazards Consideration.
10 CFR 50.91(a) states ``At the time a licensee requests an
amendment, it must provide to the Commission its analysis about the
issue of no significant hazards consideration using the standards in
Sec. 50.92.'' The following provides this analysis for the DSS-CD
long-term stability solution.
(1) Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change will implement DSS-CD as the long-term
stability solution. The DSS-CD solution is designed to identify the
power oscillation upon inception and initiate control rod insertion
to terminate the oscillations prior to any significant amplitude
growth. The DSS-CD provides protection against violation of the
Safety Limit Minimum Critical Power Ratio (SLMCPR) for anticipated
oscillations. Compliance with General Design Criteria (GDC) 10 and
12 of 10 CFR part 50, Appendix A is accomplished via an automatic
action. The DSS-CD introduces an enhanced detection algorithm that
detects the inception of power oscillations and generates an earlier
power suppression trip signal exclusively based on successive period
confirmation recognition. The existing Option III algorithms are
retained (with generic setpoints) to provide defense-in-depth
protection for unanticipated reactor instability events.
A developing instability event is suppressed by the DSS-CD
system with substantial margin to the SLMCPR and no clad damage,
with the event terminating in a scram and never developing into an
accident. In addition, the DSS-CD solution defense-in-depth features
incorporate all the
[[Page 7814]]
backup scram algorithms plus the licensed scram feature of the
existing Option III system. The DSS-CD system does not interact with
equipment whose failure could cause an accident. Scram setpoints in
the DSS-CD will be established so that analytical limits are met.
The reliability of the DSS-CD will meet or exceed that of the
existing system. No new challenges to safety-related equipment will
result from the DSS-CD solution. Because an instability event would
reliably terminate in an early scram without impact on other safety
systems, there is no significant increase in the probability of an
accident.
Proper operation of the DSS-CD system does not affect any
fission product barrier or Engineered Safety Feature. Thus, the
proposed change cannot change the consequences of any accident
previously evaluated. As stated above, the DSS-CD solution meets the
requirements of GDC 10 and 12 by automatically detecting and
suppressing design basis thermal-hydraulic oscillations prior to
exceeding the fuel SLMCPR.
Based on the above, the operation of the DSS-CD solution within
the framework of the Option III OPRM hardware will not increase the
probability or consequences of an accident previously evaluated.
(2) Will the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The DSS-CD solution operates within the existing Option III OPRM
[Oscillation Power Range Monitor] hardware. No new operating mode,
safety-related equipment lineup, accident scenario, system
interaction, or equipment failure mode was identified. Therefore,
the DSS-CD solution will not adversely affect plant equipment.
Because there are no hardware design changes * * *, there is no
change in the possibility or consequences of a failure. The worst
case failure of the equipment is a failure to initiate mitigating
action (i.e., scram), but no failure can cause an accident of a new
or different kind than any previously evaluated.
Based on the above, the proposed change to the DSS-CD solution
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
(3) Will the change involve a significant reduction in a margin
of safety?
The DSS-CD solution is designed to identify the power
oscillation upon inception and initiate control rod insertion to
terminate the oscillations prior to any significant amplitude
growth. The DSS-CD solution algorithm will maintain or increase the
margin to the SLMCPR for anticipated instability events. The safety
analyses in NEDC-33075P * * * demonstrate the margin to the SLMCPR
for postulated bounding stability events. As a result, there is no
impact on the MCPR [minimum critical power ratio] Safety Limit
identified for an instability event.
The current Option III algorithms (Period Based Detection,
Amplitude Based, and Growth Rate) are retained (with generic
setpoints) to provide defense-in-depth protection for unanticipated
reactor instability events.
Based on the above, the proposed change will not involve a
significant reduction in the margin of safety.
Conclusions: The DSS-CD stability solution has been
investigated. The BSEP licensing requirements have been evaluated
and it has been demonstrated that the DSS-CD stability solution can
be accommodated:
[sbull] Without a significant increase in the probability or
consequences of an accident previously evaluated,
[sbull] Without creating the possibility of a new or different
kind of accident from any accident previously evaluated, and
[sbull] Without exceeding any presently existing regulatory
limits or acceptance criteria applicable to the plant, which might
cause a reduction in a margin of safety.
Having made negative declarations regarding the 10 CFR 50.92
criteria, this assessment concludes that the DSS-CD stability
solution does not involve a Significant Hazards Consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602. NRC Section Chief: Allen G. Howe.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: June 7, 2002, supplemented by letter
dated January 8, 2003.
Description of amendment request: The proposed amendments would
revise the Updated Final Safety Analysis Report to eliminate credit for
the flow path from the spent fuel pool to the high pressure injection
pump as one source of primary system makeup following a tornado. The
proposed amendments would also credit the Standby Shutdown Facility as
the assured means of achieving safe shutdown for all three Oconee units
following a tornado. By letter dated January 8, 2003, Duke Energy
Corporation provided a revised No Significant Hazards Consideration
(NSHC) that supercedes the NSHC that was noticed in the Federal
Register on July 23, 2002 (67 FR 48216).
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:
Pursuant to 10 CFR 50.91, Duke Energy Corporation (Duke) has
made the determination that this amendment request involves a No
Significant Hazards Consideration by applying the standards
established by the NRC regulations in 10 CFR 50.92. This ensures
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.
The changes being requested in this amendment request involve
(1) the elimination of the Spent Fuel Pool [SFP] as a suction source
to a High Pressure Injection [HPI] pump for primary system make-up,
and (2) to fully credit the Standby Shutdown Facility (SSF) as the
primary assured means of achieving safe shutdown of all three units
following a tornado. Following the modification to fully tornado
protect the SSF, this facility becomes the station's assured flow
path for both primary make-up and secondary decay heat removal for
all three units.
Although the probability of a severe tornado strike at the
station does not change, new tornado insights gained from a review
of the current external event risk analysis have resulted in an
enhanced risk model that more accurately characterizes station
tornado damage risk. The proposed changes are part of the revised
tornado mitigation strategy that provides for an assured,
deterministic success path rather than the current strategy that is
based on risk insights and diversity for achieving safe shutdown.
This effort has resulted in an overall reduction in tornado risk at
the station and consequently, would not result in a significant
increase in the consequences of an accident previously evaluated.
Other than the fortification of walls of existing structures to
harden them against tornado damage, there are no physical changes to
the plant structures, systems, or components (SSCs), nor are there
any changes to safety limits or set points. Also, no new
radiological release pathways are created.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The changes being proposed in this amendment request do not
create the possibility of a new or different kind of accident from
any accident previously evaluated. The initial placement of the SFP-
HPI flow path into the LB [licensing basis] was based on 1989 risk
analyses that showed a potential need for primary make-up due to
inventory losses from a reactor coolant pump (RCP) seal loss-of-
cooling accident (LOCA). The upgrade of the RCP seals has
significantly reduced the probability of a seal LOCA and
subsequently, alleviated the initial reliance on the SFP-HPI flow
path for primary make-up. If multi-unit primary make-up and decay
heat removal are required following an event, the tornado protected
SSF RBMU [sic] [(RCMU) reactor coolant makeup] or SSF ASW [auxiliary
service water] pumps have the capabilities to perform these
functions for all three units.
3. Involve a significant reduction in a margin of safety.
[[Page 7815]]
As mentioned previously, new tornado insights gained from a
review of the current external event risk analysis have resulted in
an enhanced risk model that more accurately characterizes station
tornado damage risk. The proposed changes are part of the revised
tornado mitigation strategy that provides for an assured,
deterministic success path rather than a strategy that is based on
risk insights and diversity for achieving safe shutdown.
There is no safety limit, set point, or design parameter changes
required. The integrity of the fuel cladding, reactor coolant
system, and containment are preserved. Thus, 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: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC 20005.
NRC Section Chief: John A. Nakoski.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: December 30, 2002.
Description of amendment request: The proposed amendment revises
two technical specifications (TSs). The first change proposes to revise
TS 2.1.1.2, ``Minimum Critical Power Ratio Safety Limit (MCPRSL)'' to
support operation during Cycle 17 with a mixed core. The second change
proposes to revise the local power range monitor (LPRM) calibration
frequency specified in the TS for the oscillation power range monitor
(OPRM) in Surveillance Requirement (SR) 3.3.1.3.2. This change will
correct 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.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below. The licensee addresses each
change separately.
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
1. The requested change to TS 2.1.1.2, MCPRSL to support the
cycle 17 core loading does not involve any plant modifications or
operational changes that could affect system reliability,
performance, or possibility of operator error. The requested changes
do not affect any postulated accident precursors, do not affect any
accident mitigation systems, and do not introduce any new accident
initiation mechanisms. The consequences of accidents previously
evaluated are not changed because the number of rods that are
protected from transition boiling is predicted to be greater than
99.9 percent which meets the acceptance criterion in NUREG-0800,
Section 4.4.
2. The requested change to SR 3.3.1.3.2, OPRM/LPRM calibration
frequency, does not involve a modification to the plant or introduce
the probability of an operator error. The LPRMs are not the
precursor to any accident. Making the LPRM surveillance frequency
for the OPRM consistent with that approved for the RPS/APRM [reactor
protection system/average power range monitor] does not change
system reliability. The proposed LPRM surveillance frequency is
supported by the uncertainties used to perform the MCPRSL analyses.
Therefore, the number of rods that are calculated to experience
transition boiling during normal operation or anticipated
operational occurrences will not be changed and the consequences of
these events will not be increased.
Therefore, these changes do not involve a significant increase
in the probability or consequences of any 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.
1. The ATRIUM-10 fuel to be used in cycle 17 is compatible with
the co-resident SVEA-96 fuel. This compatibility is demonstrated by
application of the FRA-ANP critical power methodology to the core
design that includes the ATRIUM-10 and SVEA-96 fuel. The proposed
changes do not represent any new modes of operation, changes in
setpoints or plant modifications other than those required for the
reactor core. The change does not introduce new postulated accident
precursors or mitigation systems. Reload design and analysis will be
performed in accordance with approved NRC methodology.
2. Increasing the time interval for the OPRM/LPRM surveillance
reduces the frequency to be consistent with the LPRM surveillance
frequency for the RPS/APRM and does not involve a modification to
the plant, introduce a new operator error or revise setpoints.
Therefore, these changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
1. The proposed MCPRSL does not involve a significant reduction
in the margin of safety associated with the criterion set forth in
NUREG-0800, section 4.4. The safety limit established for the core
ensures that the criterion for the number of fuel rods allowed to
experience transition boiling will be maintained for normal plant
operation and anticipated operational transients.
The core operating limits will continue to be determined using
methodologies that have been approved by the NRC.
2. The proposed LPRM surveillance frequency is supported by the
uncertainties used to perform the MCPRSL analyses. Therefore, the
number of rods that are calculated to experience transition boiling
during normal operation or anticipated operational occurrences will
not be changed.
Therefore, implementation of the change to the MCPRSL and the
LPRM surveillance frequency 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: Thomas C. Poindexter, Esq., Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Entergy Nuclear Generation Company, Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: October 10, 2002, as supplemented on
November 22, 2002, and January 28, 2003. This notice supercedes 67 FR
68735 published on November 12, 2002, which erroneously stated that the
October 10, 2002, application was a supplement of the licensee's
application dated December 12, 2001. The October 10, 2002, replaced the
December 12, 2001, application. This notice also adds supplements dated
November 22, 2002, and January 28, 2003.
Description of amendment request: The proposed amendment would
change the Technical Specification Tables 3.2.A, 3.2.B, 4.2.A, and
4.2.B. The proposed changes affect various instrument trip level
settings and decreases the calibration frequencies for a variety of
instruments. The proposed changes also involve clarifications to the
Reactor Water Cleanup system trip configuration and the titles of
certain trip systems. In addition, the proposed changes would make
certain editorial and administrative corrections. The proposed setpoint
changes and calibration frequencies are based on the licensee's
evaluation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 7816]]
1. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The methodology used to determine the proposed trip level
settings and surveillance intervals ensure adequate performance of
the affected instrumentation. In addition, the affected instruments
are not initiators of any accident previously evaluated. Therefore,
the proposed trip level setting and surveillance intervals will not
involve a significant increase in the probability of an accident
previously evaluated.
The proposed changes to trip level settings and surveillance
intervals were establish using methodologies subject to 10 CFR
Appendix B Quality Assurance program and ensure existing
radiological limits are met. Therefore, the proposed trip level
settings and surveillance intervals will not involve a significant
increase in the consequences of an accident previously evaluated.
Other changes are editorial or administrative in nature and can
not significantly increase the probability or consequences of an
accident previously evaluated.
2. Will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
No new or different [kind] of accidents or malfunctions than
those previously analyzed in Pilgrim's UFSAR [Updated Final Safety
Analysis Report] are introduced by this proposed change because
there are no new failure modes introduced. Therefore, the proposed
changes will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. Will not involve a significant reduction in the margin of
safety.
The proposed changes to trip level settings and surveillance
intervals were established using approved methodologies subject to a
10 CFR, Appendix B, Quality Assurance program and existing
radiological limits are met. These changes do not impact Pilgrim's
configuration or operation.
Editorial and administrative type changes do not impact the
operation or configuration of Pilgrim. For the above reasons 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts, 02360-5599.
NRC Section Chief: James W. Clifford.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: December 4, 2002. This notice supercedes
68 FR 2801 published on January 21, 2003, which erroneously stated that
the December 4, 2002, application was a supplement of the licensee's
application dated May 1, 2002. The December 4, 2002, application
replaced the May 1, 2002, application in its entirety.
Description of amendment request: The proposed amendment would
extend the applicability of the current Pilgrim Nuclear Power Station
(Pilgrim) reactor pressure vessel pressure-temperature (P-T) curves
through the end of Operating Cycle (OC) 16. The current P-T curves were
approved for use in License Amendment 190, dated April 13, 2001, and
are limited to use through the end of OC 14. The proposed change would
delete the 20 and 32 Effective Full Power Year (EFPY) curves and
replace the wording of the title blocks to allow use through the end of
OC 16. The proposed amendment would change Pilgrim Technical
Specification Figures 3.6.1, 3.6.2, and 3.6.3.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change involves a request to extend the use of the
current reactor pressure vessel P-T curves for two additional OCs.
The P-T curves were generated in accordance with the fracture
toughness requirements of 10 CFR part 50, Appendix G, and American
Society of Mechanical Engineers Boiler and Pressure Vessel Code
(ASME Code), section XI, Appendix G and Regulatory Guide 1.99,
Revision 2, Radiation Embrittlement of Reactor Vessel Materials, and
were established in compliance with the methodology used to
calculate and predict effects of radiation on embrittlement of
reactor pressure vessel beltline materials. There are no physical
changes to the plant or new modes of operation being introduced by
the proposed change. Further, the proposed change does not involve a
change to any activities or equipment and is not assumed in the
safety analysis to initiate any accident sequence. The proposed
change does not adversely affect the integrity of the reactor
coolant pressure boundary such that its function in the containment
of radioactive materials is affected. Additionally, the proposed
change will not create any failure mode not bounded by previously
evaluated accidents. Therefore, the proposed change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The current P-T curves were generated in accordance with the
fracture toughness requirements of 10 CFR part 50, Appendix G, and
ASME Code, section XI, Appendix G, and were approved by the U.S.
Nuclear Regulatory Commission for use through OC 14. The proposed
change would extend use of the P-T curves for two additional OCs. No
new modes of operation are introduced by the proposed change. Plant
operation in compliance with the current P-T curves ensures
conditions in which brittle fracture of primary coolant pressure
boundary materials is avoided. Accidents involving a breach of the
primary coolant pressure boundary have previously been evaluated and
no other types of accidents associated with the proposed change have
been identified. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The proposed curves were established in compliance with the
methodology used to calculate and predict effects of radiation on
embrittlement of reactor pressure vessel beltline materials and are
estimated for 48 effective full-power years. The current curves are
approved for use through the end of OC 14 ([sim]19 EFPYs) which
provides a conservatism factor of 1.7 between the actual EFPYs at
the end of OC 14 and the end-of-life curve (32 EFPY). The change
would extend the use of the proposed curves to the end of OC 16
([sim]23 EFPYs) which provides a conservatism factor of
approximately 2.0. The actual EFPYs at the end of OC 16 is bounded
by the 48 EFPYs estimated for the current curves. 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts 02360-5599.
NRC Section Chief: James W. Clifford.
[[Page 7817]]
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of amendment request: December 12, 2002.
Description of amendment request: The proposed amendment would add
a new Surveillance Requirement (SR) to the technical specification (TS)
section 3.7.5, ``Auxiliary Feedwater (AF) System,'' which requires
operation of the diesel-driven AF pump on a monthly frequency (i.e.,
once every 31 days) for greater than or equal to 15 minutes. The
current TS SR 3.7.5.3 requires both the diesel-driven AF pump and the
motor-driven AF pump to be operated once per quarter in accordance with
the Inservice Testing Program; however, based on operating experience,
Braidwood and Byron Stations conduct the diesel-driven AF pump
surveillance on a monthly frequency to maintain a high level of
assurance that the diesel engine would automatically start when called
upon to perform its design basis 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 TS change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change adds a new TS SR to the AF System TS section
3.7.5. The new SR requires that the diesel-driven AF pump be
operated for greater than or equal to 15 minutes every month.
Operating experience has shown that conducting the diesel-driven AF
pump surveillance on a monthly frequency maintains a high level of
assurance that the diesel engine will automatically start when
called upon to perform its design basis function.
The previously analyzed events are initiated by the failure of
plant structures, systems, or components. The AF system is not
considered an initiator for any of these previously analyzed events.
The proposed change does not have a detrimental impact on the
integrity of any plant structure, system, or component that
initiates an analyzed event. No active or passive failure mechanisms
that could lead to an accident are affected. The proposed change
will not alter the operation of, or otherwise increase the failure
probability of any plant equipment that initiates an analyzed
accident. Therefore, the proposed change does not involve a
significant increase in the probability of an accident previously
evaluated.
The initial conditions of design basis accident and transient
analyses in the Byron/Braidwood Stations Updated Final Safety
Analysis Report assume the AF system is operable. The operability of
the AF system is assured by the proposed TS SR and is consistent
with the initial assumptions of the accident analyses. Since
functionality of the diesel engine can be better assured when the
diesel-driven AF pump is operated monthly vice quarterly, Exelon is
proposing to add a TS SR to operate the diesel-driven AF pump on a
monthly frequency. The proposed SR will provide higher confidence
that the diesel-driven AF pump will reliably start automatically
during an emergency condition, consistent with the AF System design
requirements, and continue to mitigate the consequences of the
associated design basis accidents. Based on this evaluation, the
proposed change does not involve a significant increase in the
consequences of an accident previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve the use or installation of
new equipment and the currently installed equipment will not be
operated in a new or different manner. No new or different system
interactions are created and no new processes are introduced. The
proposed changes will not introduce any new failure mechanisms,
malfunctions, or accident initiators not already considered in the
design and licensing bases. The current diesel-driven AF pump
surveillance procedure is already conducted on a monthly basis and
has been reviewed, approved and judged appropriate to provide high
confidence that the AF diesel engine and pump will reliably start
and operate during an emergency condition. The new SR formalizes
this monthly surveillance practice in the TS. Based on this
evaluation, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
The proposed change does not alter any existing setpoints at
which protective actions are initiated and no new setpoints or
protective actions are introduced. The design and operation of the
AF system remains unchanged and maintains the existing margins of
safety. Since the increased frequency of the diesel-driven AF pump
surveillance test maintains high assurance that the pump's diesel
engine will successfully auto-start during an emergency, the
proposed additional SR will provide high confidence that the AF
system will continue to function as designed. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety.
Based on the above, Exelon concludes that the proposed amendment
presents no significant hazards consideration under the standards
set forth in 10 CFR 50.92(c).
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.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: December 23, 2002.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) section 6, Administrative
Controls, to: (1) relocate administrative requirements discussed in
Administrative Letter 95-06 (AL 95-06), ``Relocation of Technical
Specification Administrative Controls Related to Quality Assurance,''
to the Operational Quality Assurance Program, (2) change the title of
the senior onsite official, and (3) bring the TSs into consistency with
changes in 10 CFR part 20.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes to the Seabrook Station TS do not adversely
affect accident initiators or precursors nor alter the design
assumptions, conditions, and configuration of the facility or the
manner in which the plant is operated and maintained. In addition,
the proposed changes do not affect the manner in which the plant
responds in normal operation, transient or accident conditions nor
do they change any of the procedures related to operation of the
plant. The proposed changes do not alter or prevent the ability of
structures, systems and components (SSCs) to perform their intended
function to mitigate the consequences of an initiating event within
the acceptance limits assumed in the Updated Final Safety Analysis
Report (UFSAR). The proposed changes are administrative and
editorial for the purpose of correcting or updating TS to reflect
current NRC [Nuclear Regulatory Commission] and industry
initiatives.
The proposed changes do not affect the source term, containment
isolation or radiological release assumptions used in evaluating the
radiological consequences of an accident previously evaluated in the
Seabrook Station UFSAR. Further, the proposed changes do not
increase the types
[[Page 7818]]
and amounts of radioactive effluent that may be released offsite,
nor significantly increase individual or cumulative occupational/
public radiation exposures.
Therefore, it is concluded that these proposed revisions do not
involve a significant increase in the probability or consequence of
an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed changes to the Seabrook Station TS do not change
the operation or the design basis of any plant system or component
during normal or accident conditions. The proposed changes do not
include any physical changes to the plant. In addition, the proposed
changes do not change the function or operation of plant equipment
or introduce any new failure mechanisms. The plant equipment will
continue to respond per the design and analyses and there will not
be a malfunction of a new or different type introduced by the
proposed changes.
The proposed changes are administrative in nature and only
correct, update and clarify the Seabrook Station Technical
Specifications to reflect NRC guidance, i.e., AL 95-06. The proposed
changes do not modify the facility nor do they affect the plant's
response to normal, transient or accident conditions. The changes do
not introduce a new mode of plant operation. The changes are an
enhancement and do not affect plant safety. The plant's design and
design basis are not revised and the current safety analyses remains
in effect.
Thus, these proposed revisions to the Seabrook Station TS do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed changes do not involve a significant reduction
in [a] margin of safety.
The proposed changes are administrative changes to the Seabrook
Station Technical Specifications. The safety margins established
through Limiting Conditions for Operation, Limiting Safety System
Settings and Safety Limits as specified in the Technical
Specifications are not revised nor is the plant design or its method
of operation revised by the proposed changes. Thus, it is concluded
that these proposed revisions to the Seabrook Station TS do not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis, and based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Section Chief: James W. Clifford.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: October 17, 2002.
Description of amendment request: The proposed amendment would
revise Technical Specification 3.7.9, ``Control Room Emergency
Filtration System (CREFS),'' by deleting the one-time extension to the
allowed outage time (AOT) for CREFS and the exception to the
requirements of limiting condition for operation 3.0.4 and surveillance
requirement 3.0.4 that were allowed during the AOT.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant increase in
the probability or consequences of any accident previously
evaluated.
The operability of CREFS ensures that the control room will
remain habitable for operators during and following all credible
accident conditions. The inoperability or failure of CREFS is not an
accident initiator or precursor. Therefore, the probability of an
accident previously evaluated will not be significantly increased as
a result of the proposed change. Because design limitations continue
to be met and the integrity of the reactor coolant system pressure
boundary is not challenged, the assumptions employed in the
calculation of the offsite radiological doses remain valid.
Therefore, the consequences of an accident previously evaluated will
not be significantly increased as a result of the proposed change.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a new or different kind
of accident from any accident previously evaluated.
The possibility for a new or different type of accident from any
accident previously evaluated is not created as a result of this
amendment. The evaluation of the effects of the proposed changes
indicate that all design standards and applicable safety criteria
limits are met. These changes therefore do not cause the initiation
of any new or different accident nor create any new failure
mechanisms.
Equipment important to safety will continue to operate as
designed.
Additionally, the changes do not result in any event previously
deemed incredible being made credible. The changes also do not
result in more adverse conditions or result in any increase in the
challenges to safety systems. Therefore, operation of the Point
Beach Nuclear Plant in accordance with the proposed amendments will
not create the possibility of a new or different type of accident
from any accident previously evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant reduction
in a margin of safety.
There are no new or significant changes to the initial
conditions contributing to accident severity or consequences. The
proposed amendment will not otherwise affect the plant protective
boundaries, will not cause a release of fission products to the
public, nor will it degrade the performance of any other structures,
systems or components (SSCs) important to safety. Therefore,
deleting the one-time extension to the CREFS AOT will 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
amendment request involves no significant hazards consideration.
Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: L. Raghavan.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: June 28, 2002, as supplemented on
December 18, 2002, and January 18, 2003.
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TSs) by relaxing the secondary
containment requirements and eliminating the Filtration, Ventilation,
and Recirculation System (FRVS) charcoal filters.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the change involve a significant increase in the
probability or consequences of an accident previously analyzed?
Response: No.
The definition of CORE ALTERATIONS has been revised to define
that control rod movement, provided there are no fuel assemblies in
the associated core cell, is not a core alteration. This is
consistent with Standard Technical Specifications (STS) NUREG-1433
Vol.1, Rev. 2, Standard Technical Specifications, General Electric
Plants, BWR/4 [Boiling Water Reactor, Type 4].
The TS presently provide a period of 7 days to restore an
inoperable FRVS ventilation unit when performing activities with the
potential for draining the reactor vessel or discontinue such
activities.
[[Page 7819]]
Operation of the redundant train will ensure that the remaining
subsystem is operable, that no failures, which could prevent
automatic actuation, have occurred and that any other failures will
be readily detected. This is consistent with STS, NUREG-1433 Vol.1,
Rev. 2, Standard Technical Specifications, General Electric Plants,
BWR/4.
The proposed changes associated with the FHA [fuel-handling
accident] do not involve a change to structures, components, or
systems that would affect the probability of an accident previously
evaluated in the Hope Creek Updated Final Safety Analysis Report
(UFSAR). The FHA for the HCGS [Hope Creek Generating Station] is
defined as a drop of a fuel assembly over irradiated assemblies in
the reactor core 24 hours after reactor shutdown. AST [accident
source term] is used to evaluate the dose consequences of a
postulated accident. The FHA has been analyzed without credit for
Secondary Containment, Filtration Recirculation and Ventilation
System (FRVS), and Control Room Emergency Filtration (CREF) system.
The resultant radiological consequences are within the acceptance
criteria set forth in 10 CFR 50.67 and Regulatory Guide 1.183. This
amendment does not alter the methodology or equipment used directly
in fuel handling operations. The equipment hatch, the personnel air
locks, nor any other containment penetration, nor any component
thereof is an accident initiator. Actual fuel handling operations
are not affected by the proposed changes. Therefore, the probability
of a Fuel Handling Accident is not affected with the proposed
amendment. No other accident initiator is affected by the proposed
changes.
The Loss of Coolant Accident (LOCA) Dose Calculation has been
revised to (1) eliminate credit for the FRVS recirculation charcoal
filters, (2) reduce credited efficiency of FRVS vent charcoal
filters, (3) reduce Engineered Safety Feature (ESF) leakage from 10
gpm to 1 gpm and (4) reduce control room unfiltered in-leakage to
350 cfm [cubic feet per minute]. These proposed changes do not
eliminate any safety system. The changes are only associated with
the credit provided by the system in reducing the radiological
consequences and therefore, do not affect any accident initiator.
The results of that analysis show that the Exclusion Area Boundary
(EAB), Low Population Zone (LPZ), and Control Room (CR) doses are of
the same order of magnitude as the previous analysis and remain
within the acceptance criteria in 10 CFR 50.67 and Regulatory Guide
1.183.
Therefore, this proposed amendment does not involve a
significant increase in the probability of occurrence or
consequences of an accident previously analyzed.
(2) Does the change create the possibility of a new or different
kind of accident from any accident previously analyzed?
Response: No.
The proposed amendment will not create the possibility for a new
or different type of accident from any accident previously
evaluated. Changes to the allowable activity in the primary and
secondary systems do not result in changes to the design or
operation of these systems. The evaluation of the effects of the
proposed changes indicates that all design standard and applicable
safety criteria limits are met.
Equipment important to safety will continue to operate as
designed. Component integrity is not challenged. The changes do not
result in any event previously deemed incredible being made
credible. The changes do not result in more adverse conditions or
result in any increase in the challenges to safety systems. The
systems affected by the changes are used to mitigate the
consequences of an accident that has already occurred. The proposed
TS changes and modifications do not significantly affect the
mitigative function of these systems.
Therefore, this proposed amendment does not create the
possibility of a new or different kind of accident from any
previously analyzed.
(3) Does the change involve a significant reduction in [a]
margin of safety?
Response: No.
The proposed changes revise the TS to establish operational
conditions where specific activities represent situations during
which significant radioactive releases can be postulated. These
operational conditions are consistent with the design basis analysis
and are established such that the radiological consequences are at
or below the regulatory guidelines. Safety margins and analytical
conservatisms are retained to ensure that the analysis adequately
bounds all postulated event scenarios. The proposed TS continue to
ensure that the TEDE [total effective dose equivalent] for the CR,
the EAB, and LPZ are below the corresponding acceptance criteria
specified in 10 CFR 50.67 and RG1.183.
Therefore, these changes do not involve a significant reduction
in [a] margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James Clifford.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: October 9, 2002, as supplemented
November 22, 2002, and December 6, 2002.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 6.8.4.f, ``Primary Containment
Leakage Rate Testing Program,'' to allow a one-time interval extension
to the requirement.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously analyzed?
Response: No.
The proposed revision to section 6.8.4.f adds a one-time
extension to the current interval for containment integrated leak
rate test (ILRT). The current test interval of 10 years, based upon
past performance, would be extended on a one-time basis to 15 years
from the last ILRT. The proposed extension to ILRT testing cannot
increase the probability of an accident previously evaluated since
the containment ILRT testing extension is not a modification to
plant systems, nor a change to plant operation that could initiate
an accident. The proposed extension to Type A testing does not
involve a significant increase in the consequences of an accident
since research documented in NUREG-1493, ``Performance-Based
Containment Leak-Test Program,'' found that very few potential
containment leakage paths fail to be identified by Type B and C
tests. The NUREG concluded that reducing the ILRT testing frequency
to once per twenty years would lead to an imperceptible increase in
risk. Containment performance monitoring is performed in accordance
with the Maintenance Rule (10 CFR 50.65) and inspections required by
American Society of Mechanical Engineers (ASME) code are performed
in order to identify indications of containment degradation that
could affect leak tightness. Type B and C testing required by the
technical specifications (TS) will identify any containment opening,
such as valves, that would otherwise be detected by the ILRT. Reg.
Guide 1.174 provides guidance for determining the risk impact of
plant-specific changes to the licensing basis. It also recommends
the use of risk analysis techniques to ensure and show that the
proposed change is consistent with the defense-in-depth philosophy.
The increase in large early release frequency (LERF) resulting from
a change in the ILRT test frequency from the current once in every
10 years to once in every 15 years is less than 1E-7 per year,
thereby meeting Regulatory Guide 1.174 definition of a very small
change in risk. The change in conditional containment failure
probability (CCFP) is estimated to be 0.25% for the proposed change.
These factors show that an ILRT test extension will not represent a
significant increase in the consequences of an accident.
Therefore, this proposed amendment does not involve a
significant increase in the probability of occurrence or
consequences of an accident previously analyzed.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously analyzed?
Response: No.
The proposed revision to section 6.8.4.f adds a one-time
exception to the current interval for the ILRT. The current test
interval of 10 years, based upon past performance, would be extended
on a one-time basis to 15 years from the last Type A test. Primary
containment is designed to contain energy and fission products
during and after an event. The Individual Plant
[[Page 7820]]
Examination (IPE) identifies events that lead to containment
failure. Revision to the ILRT test interval does not change this
list of events. There are no physical changes being made to the
plant and there are no changes to the operation of the plant that
could introduce a new failure mode creating a new or different kind
of accident.
Therefore, this proposed amendment does not create the
possibility of a new or different kind of accident from any
previously analyzed.
3. Does the change involve a significant reduction in the margin
of safety?
Response: No.
The proposed revision to section 6.8.4.f adds a one-time
extension to the current interval for the ILRT. The current test
interval of 10 years, based upon past performance, would be extended
on a one-time basis to 15 years from the last ILRT. The proposed
extension to ILRT testing interval will not significantly reduce the
margin of safety. The NUREG-1493 generic study of the effects of
extending containment leakage testing found that a 20-year exception
in ILRT leakage testing resulted in an imperceptible increase in
risk to the public. NUREG-1493 found that the containment leakage
rate contributes a very small amount to the individual risk, and
that the decrease in Type A testing frequency would have a minimal
affect on this risk since most potential leakage paths are detected
by Type C testing. Type B and Type C testing will continue to be
performed at a frequency currently required by the Technical
Specifications (TS). The containment inspections being performed in
accordance with ASME, section XI, and Maintenance Rule (10 CFR
50.65) provide a high degree of assurance that the containment will
not degrade in a manner that is only detectable by Type A testing.
Reg. Guide 1.174 provides guidance for determining the risk
impact of plant-specific changes to the licensing basis. It also
recommends the use of risk analysis techniques to ensure and show
that the proposed change is consistent with the defense-in-depth
philosophy. The increase in large early release fraction (LERF)
resulting from a change in the ILRT test frequency from the current
once in every 10 years to once in every 15 years is less than 1E-7
per year, thereby meeting Regulatory Guide 1.174 definition of a
very small change in risk. The change in conditional containment
failure probability (CCFP) is estimated to be 0.25% for the proposed
change.
Therefore, these changes do not involve a significant reduction
in 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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James Clifford.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: July 25, 2002, as supplemented on
October 21, 2002.
Description of amendment request: The proposed change would revise
Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2, Technical
Specifications (TSs) Surveillance Requirement (SR) 4.0.3 to extend the
delay period, before entering a Limiting Condition for Operation,
following a missed surveillance. The delay period would be extended
from the current limit of up to 24 hours, to ``* * * up to 24 hours or
up to the limit of the specified frequency, whichever is greater.'' In
addition, the following requirement would be added to SR 4.0.3: ``A
risk evaluation shall be performed for any surveillance delayed greater
than 24 hours and the risk impact shall be managed.'' PSEG is also
proposing changes to adopt a TS Bases Control Program and changes to SR
4.0.1.
The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice
of opportunity for comment in the Federal Register on June 14, 2001 (66
FR 32400), on possible amendments concerning missed surveillances,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process (CLIIP). The NRC staff subsequently issued a notice
of availability of the models for referencing in license amendment
applications in the Federal Register on September 28, 2001 (66 FR
49714). The licensee affirmed the applicability of the model NSHC
determination for amendments concerning missed surveillances in its
original application dated July 25, 2002. The proposed amendment would
also make administrative changes to SRs 4.0.1 and 4.0.3 to be
consistent with NUREG-1431, Revision 2, ``Standard Technical
Specifications, Westinghouse Plants.'' These changes are necessary to
make the current Salem TSs compatible with the proposed CLIIP changes
for missed surveillances.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
[Specification 4.0.3]
The proposed change relaxes the time allowed to perform a missed
Surveillance. The time between Surveillances is not an initiator to
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.
[Specification 4.0.1]
The proposed additional requirement equating failure to meet a
surveillance with failure to meet the [limiting condition for
operation] is consistent with current interpretation of the
technical specifications. This change, along with relocation and
rewording of existing requirements from Specification 4.0.3, are
administrative in nature and do not adversely affect accident
initiators, design functions, facility configuration or the manner
of operation or control. The ability of structures, systems and
components to perform their intended function remains unaffected.
[Bases Control Program]
The proposed change to adopt a Technical Specification Bases
Control Program is also administrative in nature and does not
adversely affect accident initiators, design functions, facility
configuration or the manner of operation or control. The ability of
structures, systems or components to perform their intended function
remains unaffected. Future changes to the TS Bases will continue to
be administratively controlled in accordance with the requirements
of 10 CFR 50.59.
Therefore, these three changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
None of the three proposed changes involves a physical
alteration of the plant (no new or different type of equipment will
be installed) or a change in the methods governing normal plant
operation. Thus, these changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
[Specification 4.0.3]
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 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
[[Page 7821]]
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.
[Specification 4.0.1]
The proposed changes to TS 4.0.1, including relocation and
rewording of existing requirements from Specification 4.0.3, are
administrative in nature and do not reduce the level of programmatic
or procedural controls associated with the Surveillance
Requirements. There are no substantive differences in meaning or
intent between the existing specifications and the corresponding STS
requirements. Further, these changes have no impact on equipment
design, configuration, analytical basis, setpoints or operation.
[Bases Control Program]
The proposed change to adopt a Technical Specification Bases
Control Program is also administrative in nature and does not reduce
the level of programmatic or procedural controls associated with the
Bases. There is no impact on equipment design, configuration,
analytical basis, setpoints or operation.
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 and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: December 19, 2002.
Description of amendment request: The proposed amendments would
change the Operating Licenses and Technical Specifications associated
with an increase in the licensed reactor power level of 1.5 percent for
each reactor (from 2763 megawatts thermal (MWt) to 2804 MWt).
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:
[Southern Nuclear Company] SNC's conclusion that the proposed
change to the Plant Hatch Unit 1 and 2 Operating Licenses and
Technical Specifications does not involve a significant hazards
consideration is based upon the following:
1. The proposed amendment does not change involve a significant
increase in the probability or consequences of an accident
previously evaluated[.]
The comprehensive analytical efforts performed to support the
proposed uprate conditions included a review and evaluation of all
components and systems that could be affected by this change.
Performance requirements for these systems were evaluated and found
acceptable. Furthermore, evaluation of accident analyses confirmed
the effects of the proposed uprate are bounded by the current dose
analyses. The systems will function as designed. The performance
requirements for these systems were evaluated and found acceptable.
The primary loop components (e.g., reactor vessel, reactor
internals, control rod drive housings, piping and supports, and
recirculation pumps) continue to comply with their applicable
structural limits and will continue to perform their intended design
functions. Thus, the probability of a structural failure of these
components is not increased as a result of this change.
The Nuclear Steam Supply System (NSSS) systems will still
perform their intended design functions during normal and accident
conditions. The balance-of-plant (BOP) systems and components will
continue to meet their applicable structural limits and perform
their intended design functions. Thus, the probability of a
structural failure of these components is not increased as a result
of this change.
The NSSS/BOP interface systems will continue to perform their
intended design functions. The safety relief valves and containment
isolation valves still meet design sizing requirements at the
uprated power level.
Because the integrity of the plant will not be affected by
operation at the uprated condition, SNC concluded that all
structures, systems, and components required to mitigate a transient
remain capable of fulfilling their intended functions. The reduced
uncertainty in the flow input to the core thermal power uncertainty
measurement allows most of the current safety analyses to be used,
with small changes to the core operating limits, to support
operation at a core power of 2804 MWt. Other analyses performed at a
nominal power level were either evaluated or reperformed for the
1.5% increased power level. The results demonstrate that the
applicable analysis acceptance criteria continue to be met at the
1.5% uprate conditions. Thus, all Plant Hatch Final Safety Analysis
Report accident analyses continue to demonstrate compliance with the
relevant event acceptance criteria. The analyses performed to assess
the effects of mass and energy release remain valid. The source
terms used to assess radiological consequences were reviewed and
determined to bound operation at the 1.5% uprated condition.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment will not change create the possibility
of a new or different kind of accident from any previously
evaluated.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed change.
All systems, structures, and components previously required for the
mitigation of a transient remain capable of fulfilling their
intended design functions. The proposed change will have no adverse
effect on any safety-related system or component and does not
challenge the performance or integrity of any safety-related system.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. The proposed amendment will not involve a significant
reduction in a margin of safety.
Operation at the uprated power condition does not involve a
significant reduction in a margin of safety. Analyses of the primary
fission product barriers confirm that all relevant design criteria
remain satisfied, both from the standpoint of the integrity of the
primary fission product barrier and from the standpoint of
compliance with the required acceptance criteria. As appropriate,
all evaluations were performed using methods that were either
reviewed and approved by the NRC, or are in compliance with
regulatory review guidance and standards. Therefore, the proposed
change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: John A. Nakoski.
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: November 14, 2002.
Description of amendment request: The proposed amendment would
delete
[[Page 7822]]
the turbine missile design basis from the Updated Final Safety Analysis
Report (UFSAR).
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 turbine missile generation probability will not be
significantly increased by elimination of the regulatory commitments
in the UFSAR. No plant changes are proposed that would significantly
increase the probability of turbine missile generation. Turbine
missile generation does not pose a credible threat to safety related
components and consequently has no potential to increase
radiological consequences.
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 no physical modification of the
plant or different operating configurations.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Turbine missiles do not constitute a credible threat to nuclear
safety at STP [South Texas Project]. They are not a consideration in
any plant safety analysis. Changing the regulatory commitment with
regard to design for turbine missiles has no effect on any margin of
safety.
Based upon the analysis provided herein, the proposed amendments
do not involve a significant hazards consideration.
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, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: January 14, 2003 (TS 02-08).
Description of amendment request: The proposed amendment would
revise applicability requirements for TS 3.3.9.4, ``Containment
Building Penetrations.'' This revision will modify the current
applicability requirement associated with movement of ``irradiated
fuel'' by adding a new applicability statement for the containment
building equipment door.
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?
The proposed change revises the applicability of the containment
building penetration function and associated action. This change
does not alter the function of the penetrations but does revise when
the feature is required to be available for the mitigation of
postulated accidents. These penetrations only function to minimize
the release of radioactive material for accident mitigation and are
not considered to be a source of any postulated accident. The
analysis verifies that a fuel handling accident (FHA) occurring at
least 100 hours after being critical in a reactor core will not
result in dose consequences above the regulatory limits without the
containment closure function provided by the CBED [containment
building equipment door]. The applicability and action for the CBED
will not be changed when movement of recently irradiated fuel is in
progress and this function ensures acceptable dose consequences.
Therefore, the proposed change will not increase the probability of
an accident because the penetration function has not been altered
and this function is not a potential source for accidents.
Additionally, the proposed change will not significantly increase
the consequences of an accident because the analysis has verified
that dose consequences will be maintained less than the required
regulatory limits.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change only modifies when containment building
penetrations need to be available for accident mitigation and does
not alter their function, design, or operation. These penetrations
only serve to minimize the release of radioactive material in the
event of postulated accidents and do not have the potential to
create an accident. Since the function of the penetrations is not
being changed and they do not have an accident generation potential,
the possibility of a new or different kind of accident is not
created.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The proposed change will not alter the function, design, or
operation of the containment building penetrations for postulated
accidents that require this feature for the mitigation of the event.
The analysis has determined that the CBED availability can be
limited to those activities that involve the movement of irradiated
fuel that has been in a critical reactor core within the previous
100 hours. Therefore, not requiring the CBED to be available 100
hours or longer afterwards will not impact plant safety or result in
dose consequences above established regulatory limits. The proposed
change will not alter any setpoints or other functions that serve to
maintain the safety limits. Therefore, the proposed change will not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Yankee Atomic Electric Co., Docket No. 50-29, Yankee Nuclear Power
Station (YNPS) Franklin County, Massachusetts
Date of amendment request: January 14, 2003.
Description of amendment request: The proposed amendment will
revise the Yankee Rowe Nuclear Power Station License and Technical
Specifications to delete operational and administrative requirements
that would no longer be required once the spent nuclear fuel has been
transferred from the spent fuel pool to the Independent Spent Fuel
Storage Installation (ISFSI).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
No. The proposed changes reflect the complete transfer of all
spent nuclear fuel from the Spent Fuel Pit to the Independent Spent
Fuel Storage Installation (ISFSI). Design basis accidents related to
the Spent Fuel Pit are discussed in the YNPS FSAR. These postulated
accidents are predicated on spent nuclear fuel being stored in the
Spent Fuel Pit. With the removal of the spent fuel from the Spent
Fuel Pit, there are no remaining important to safety systems
required to be monitored and there are no remaining credible
accidents that require that actions of a Certified Fuel Handler or
non-Certified Fuel Handler to prevent occurrence or mitigate the
consequences.
The YNPS FSAR provides a discussion of radiological events
postulated to occur as a result of decommissioning with the bounding
consequence resulting from a materials handling event. The proposed
changes do not have an adverse impact on decommissioning activities
or any of their postulated consequences.
[[Page 7823]]
The proposed change to the Design Features section of the
Technical Specifications clarifies that the spent fuel is being
stored in dry casks within an ISFSI. The probability or consequences
of accidents at the ISFSI are evaluated in the dry cask vendor's
FSAR and are independent of the accidents evaluated in the YNPS
FSAR.
Based on the above, the proposed changes will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Does the proposed license amendment create the possibility of a
new or different kind of accident from any accident previously
evaluated?
No. The proposed changes reflect the reduced operational risks
as a result of the spent nuclear fuel being transferred to dry casks
within an ISFSI. The proposed changes do not modify any physical
systems, or components. The plant conditions for which the YNPS FSAR
design basis accidents relating to spent fuel have been evaluated
are no longer applicable. The aforementioned proposed changes do not
affect any of the parameters or conditions that could contribute to
the initiation of an accident. Design basis accidents associated
with the dry cask storage of spent fuel are already considered in
the dry cask system's Final Safety Analysis Report. No new accident
scenarios are created as a result of deleting non-applicable
operational and administrative requirements. Therefore, the proposed
changes will not create the possibility of a new or different kind
of accident from any previously evaluated.
Does the proposed license amendment involve a significant
reduction in a margin of safety?
No. As described above, the proposed changes reflect the reduced
operational risks as a result of the spent nuclear fuel being
transferred to dry casks within an ISFSI. The design basis and
accident assumptions within the YNPS FSAR and the Defueled Technical
Specifications relating to spent fuel are no longer applicable. The
proposed changes do not affect remaining plant operations, systems,
or components supporting decommissioning activities. In addition,
the proposed changes do not result in a change in initial
conditions, system response time, or in any other parameter
affecting the course of a decommissioning activity accident
analysis. Therefore, the proposed changes will not involve a
significant reduction in the margin of safety.
Based on the considerations noted above, it is concluded that
the proposed changes will not endanger the public health and 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: Thomas Dignan, Esquire, Ropes and Gray, One
International Place, Boston, Massachusetts 02110-2624.
NRC Section Chief: Scott W. Moore.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: January 16, 2003.
Brief description of amendment request: The proposed amendment
would revise the applicable Technical Specifications requirements for
rod position monitoring during the current operating cycle (Cycle 22)
to allow the use of an alternate method of determining rod position.
This would be effective until repair of the indication system can be
completed during the next shutdown of sufficient duration.
Date of publication of individual notice in Federal Register:
January 24, 2003 (68 FR 3566).
Expiration date of individual notice: February 7, 2003, for
comments; February 24, 2003, for hearings.
Florida Power and Light Company, Docket No. 50-251, Turkey Point Plant,
Units 3 and 4, Miami-Dade County, Florida
Date of application for amendments: November 26, 2002.
Brief description of amendments: The proposed license amendments
would revise Technical Specifications (TSs) to increase the total spent
fuel wet storage capacity by adding a spent fuel storage rack in the
cask area in each unit's spent fuel pool. Also, it would revise the
location called out in the Design Features sections 5.6.1.1a and b of
the TSs referring to Updated Final Safety Analysis Report Appendix 14D,
rather than referring to Westinghouse Report WCAP-14416-P.
Date of publication of individual notice in the Federal Register:
January 28, 2003 (68 FR 4246).
Expiration date of individual notice: February 27, 2003.
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].
[[Page 7824]]
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant, Middlesex County, Connecticut
Date of amendment request: May 29, 2001, and its supplements dated
August 29, 2001, and September 24, 2002.
Brief description of amendment: The amendment revises paragraph
2.C.(5), ``Physical Protection,'' of Facility Operating License No.
DPR-61 to reference the Defueled Physical Security Plan that includes
the security plan for the Independent Spent Fuel Storage Installation.
Date of issuance: January 30, 2003.
Effective date: January 30, 2003, and shall be implemented within
30 days from the date of issuance and prior to the transfer of spent
nuclear fuel to the Independent Spent Fuel Storage Installation.
Amendment No.: 199.
Facility Operating License No. DPR-61: The amendment revised the
Operating License.
Date of initial notice in Federal Register: August 22, 2001 (66 FR
44163). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 30, 2003.
No significant hazards consideration comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: August 8, 2002, as supplemented
October 23, 2002.
Brief description of amendment: The amendment authorized changes to
the Updated Final Safety Analysis Report (USFAR) for Fermi 2 by
allowing implementation of the Boiling Water Reactor Vessel and
Internals Project reactor pressure vessel Integrated Surveillance
Program as the basis for demonstrating the compliance with the
requirements of Appendix H, ``Reactor Vessel Material Surveillance
Program Requirements,'' to 10 CFR part 50.
Date of issuance: January 30, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 152.
Facility Operating License No. NPF-43: Amendment authorizes changes
to the USFAR.
Date of initial notice in Federal Register: September 3, 2002 (67
FR 56320). The October 23, 2002, supplemental letter provided
additional clarifying information that did not change the original no
significant hazards consideration determination or expand the amendment
beyond the scope of the original notice. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
January 30, 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: April 18, 2002, as supplemented
August 7, and October 9 and October 30, 2002, and January 15, 2003.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3.7.15 in response to Boraflex degradation to
provide revised spent fuel pool (SFP) storage criteria, and revised
fuel enrichment and burnup requirements that take credit for soluble
boron. TS 4.3.1 is revised to increase the required soluble boron
credit from a concentration of 730 parts per million (ppm) to 850 ppm
to ensure acceptable levels of subcriticality in the SFPs. Associated
changes to the TS Bases are also included.
Date of issuance: February 4, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 210 & 191.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 25, 2002 (67 FR
42820). The supplements dated August 7, and October 9 and October 30,
2002, and January 15, 2003, provided clarifying information that did
not change the scope of the April 18, 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 February 4, 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: October 22, 2002.
Brief description of amendment: The amendment deletes TS 5.5.3,
``Post Accident Sampling System (PASS),'' and thereby eliminates the
requirements to have and maintain the PASS at Columbia Generating
Station. The amendment also addresses related changes to TS 5.5.2,
``Primary Coolant Sources Outside Containment,'' and License Condition
2.C.(13), ``Post Accident Sampling.''
Date of issuance: January 27, 2003.
Effective date: January 27, 2003, to be implemented within 60 days
from the date of issuance.
Amendment No.: 184.
Facility Operating License No. NPF-21: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: December 24, 2002 (67
FR 78518). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 27, 2003.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: May 14, 2002, as supplemented by letters
dated July 9, August 2, September 16, and November 7 and 22, 2002.
Brief description of amendment: This amendment increases the
licensed power level by approximately 1.7 percent from 3,039 megawatts
thermal (MWt) to 3,091 MWt. These changes result from increased
feedwater flow measurement accuracy to be achieved by utilizing high
accuracy ultrasonic flow measurement instrumentation.
Date of issuance: January 31, 2003.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 129.
Facility Operating License No. NPF-47: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: June 11, 2002 (67 FR
40022). The July 9, August 2, September 16, and November 7 and 22,
2002, 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 amendment is contained in a
Safety Evaluation dated January 31, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: April 24, 2001, as supplemented
on May 22, 2002.
Brief description of amendment: The amendment revised information
in the
[[Page 7825]]
Final Safety Analysis Report regarding the protection of the component
cooling water (CCW) system from natural phenomena. The change addresses
the fact that a portion of one safety-related loop of the CCW system is
routed through the fuel storage building, where the structure was not
designed to protect the CCW piping from the effects of natural
phenomena.
Date of issuance: January 27, 2003.
Effective date: January 27, 2003.
Amendment No.: 214.
Facility Operating License No. DPR-64: Amendment revised the Final
Safety Analysis Report.
Date of initial notice in Federal Register: October 3, 2001 (66 FR
50466). The May 22, 2002, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated January 27, 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: February 26, 2002, as revised by
letters dated October 9 and 30, 2002.
Brief description of amendment: The amendment revises the
definition of Operable in Technical Specification (TS) 1.0.K with
respect to support system requirements for alternating current power
sources. Conforming changes are also made to a specific support system
TS in Sections 3/4.5, ``Core and Containment Cooling Systems'', 3/4.7,
``Station Containment Systems'', and 3/4.10, ``Auxiliary Electrical
Power Systems,'' and associated Bases.
Date of Issuance: February 4, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 90 days.
Amendment No.: 213.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: (67 FR 78519). The
Commission's related evaluation of this amendment is contained in a
Safety Evaluation dated February 4, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: September 19, 2002, as
supplemented December 26, 2002.
Brief description of amendments: The amendments add a new
analytical method to Technical Specifications (TS) section 5.6.5,
``Core Operating Limits Report.'' The change supports the core design
efforts used for the Unit 2 refueling outage which began on January 21,
2003.
Date of issuance: February 4, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 159 & 145.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 15, 2002 (67 FR
63694). The December 26, 2002, supplemental letter provided clarifying
information that was within the scope of the initial notice and did not
change the initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 4, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: May 31, 2002, as supplemented
by letter dated October 16, 2002.
Brief description of amendments: These amendments revised Technical
Specifications (TSs) 3.8.2.1, ``DC Sources--Operating,'' and 3.8.2.2,
``DC Sources--Shutdown''; and added the new Specification 6.8.4.i,
``Battery Monitoring and Maintenance Program.'' The changes also
included the relocation of the following TS items to a licensee-
controlled program: (1) A number of surveillance requirements that
require the performance of preventive maintenance, and (2) certain
battery and battery cell parameter values that are periodically
verified to monitor early indications of DC subsystem degradation.
Date of issuance: January 29, 2003.
Effective date: As of date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 164 and 126.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 17, 2002 (67
FR 58643). The supplement dated October 16, 2002, 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 amendments is contained in a
Safety Evaluation dated January 29, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station Units 1 and 2, Lake County, Illinois
Date of application for amendments: February 28, 2001, as
supplemented by letter dated June 13, 2002.
Brief description of amendments: Revise the Technical
Specifications to eliminate the requirement for at least one person
qualified to stand watch to be present in the control room when nuclear
fuel is stored in the spent fuel pool.
Date of issuance: January 31, 2003.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: 183 and 170.
Facility Operating License Nos. DPR-39 and DPR-48: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 27, 2001 (66 FR
34283).
The June 13, 2002, supplemental letter 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 January 31, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334,
Beaver Valley Power Station, Unit No. 1, Beaver County, Pennsylvania
Date of application for amendment: May 31, 2002, as supplemented
July 19, and September 3, 2002.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.1.1.4, upper limit for the moderator temperature
coefficient (MTC), from 0 x 10-4 change in reactivity per
degree Fahrenheit ([Delta]k/k/[deg]F) to +0.2 x 10-4
[Delta]k/k/[deg]F for power
[[Page 7826]]
levels up to 70 percent of rated thermal power (RTP), and ramping
linearly to 0 x 10-4 [Delta]k/k/[deg]F from 70 percent to
100 percent RTP. The change is needed to address future core designs
with higher energy requirements, associated with plant operation at
higher capacity factors.
Date of issuance: February 6, 2003.
Effective date: As of date of issuance and shall be implemented
within 60 days.
Amendment No.: 251.
Facility Operating License No. DPR-66: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 17, 2002 (67
FR 58644). The July 19, and September 3, 2002, letters provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated February 6, 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: March 27, 2002, as supplemented
on October 7, 2002.
Brief description of amendment: The amendment revises the Technical
Specifications section 3.6.3, ``Emergency Power Sources,'' to extend
the current allowable outage time for an inoperable diesel generator
from 7 days to 14 days, and section 3.4.4, ``Emergency Ventilation
System,'' and section 3.4.5, ``Control Room Air Treatment System,'' to
reflect the change to section 3.6.3.
Date of issuance: February 3, 2003.
Effective date: February 3, 2003.
Amendment No.: 179.
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 30, 2002 (67 FR
21290). The October 7, 2002, 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 the amendment is
contained in a Safety Evaluation dated February 3, 2003.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: April 25, 2002. The application
was initially submitted to the Nuclear Regulatory Commission with an
incorrect date of April 25, 2001. The Nuclear Management Company, LLC,
subsequently submitted a letter dated May 30, 2002, correcting the date
of the application as April 25, 2002.
Brief description of amendment: The amendment changes Technical
Specification (TS) 3.7/4.7, ``Containment Systems,'' to allow the use
of 10 CFR part 50, Appendix J, Option B, for Types B and C containment
leak rate testing and adds a new TS section 6.8.M, ``Programs and
Manuals--Primary Containment Leakage Rate Testing Program.''
Date of issuance: February 4, 2003.
Effective date: As of the date of issuance and to be implemented
within 75 days.
Amendment No.: 132.
Facility Operating License No. DPR-22: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 3, 2002 (67
FR 56325).
The May 30, 2002, letter corrected the date of the application and
did not change the NRC staff's initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated February 4, 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: July 25, 2002, as supplemented
by letter dated October 23, 2002.
Brief description of amendments: These amendments revise the
Susquehanna Steam Electric Station Final Safety Analysis Report (SSES
FSAR) by replacing the current plant-specific reactor pressure vessel
material surveillance program with the Boiling Water Reactor Integrated
Surveillance Program.
Date of issuance: February 6, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 208 and 182.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the SSES FSAR.
Date of initial notice in Federal Register: September 3, 2002 (67
FR 56328). The October 23, 2002, supplemental letter provided
additional information that clarified the application, but 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 February 6, 2003.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: October 23, 2002.
Brief description of amendment: The amendment updates the reference
to 10 CFR 20.203 with the corresponding reference to 10 CFR 20.1601.
Hope Creek Generating Station Technical Specification (TS) 6.12, ``High
Radiation Area,'' is revised to be consistent with the Standard TSs,
General Electric Plants (NUREG-1433, Rev. 2).
Date of issuance: January 30, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 142.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75884).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 30, 2003.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: August 21, 2002.
Brief description of amendments: The amendments revise the
technical specifications (TSs) to replace reference to specific valves
for preventing uncontrolled boron dilution. The revised TSs incorporate
a general statement for preventing uncontrolled boron dilution,
consistent with the improved standard TSs.
Date of issuance: January 27, 2003.
Effective date: January 27, 2003.
Amendment Nos.: Unit 1-149; Unit 2-137.
[[Page 7827]]
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the TSs.
Date of initial notice in Federal Register: October 1, 2002 (67 FR
61686).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 27, 2003.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: May 23, 2002.
Brief description of amendments: The amendments relocated the
shutdown margin limits to the Core Operating Limits Report and modified
certain boration requirements consistent with NUREG-1431. The
amendments also correct some typographical errors in the Technical
Specification pages.
Date of issuance: February 4, 2003.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: Unit 1-150; Unit 2-138.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 25, 2002 (67 FR
42830).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 4, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant, Units 2 and 3, Limestone County, Alabama
Date of application for amendments: November 6, 2002.
Brief description of amendments: The amendments revised the Browns
Ferry Nuclear Plant, Units 2 and 3, Updated Final Safety Analysis
Report (UFSAR) to modify the basis for TVA's compliance with the
requirements of Appendix H to title 10 of the Code of Federal
Regulations part 50, ``Reactor Vessel Material Surveillance Program
Requirements.''
Date of issuance: January 28, 2003.
Effective date: As of the date of issuance, to be incorporated into
the UFSAR at the time of its next update.
Amendment Nos.: 279 & 238.
Facility Operating License Nos. DPR-52 and DPR-68: Amendments
revised the UFSAR.
Date of initial notice in Federal Register: November 26, 2002 (67
FR 70770). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated January 28, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: September 3, 2002, as
supplemented October 17, 2002, and January 29, 2003.
Brief description of amendments: The amendments revise Technical
Specification (TS) Surveillance Requirement (SR) 4.0.3 to extend the
delay period, before entering a Limiting Condition for Operation,
following a missed surveillance. Thise changes to SR 4.0.3 will allow
an extension of up to 24 hours or the limit of the surveillance
frequency, whichever is greater. The amendments also include editorial
changes to make the revised TS consistent with the Standard TS for
Westinghouse plants. In addition, the amendments include the adoption
of the TS Bases Control Program listed in NUREG-1431, Revision 2.
Date of issuance: February 5, 2003.
Effective date: As of the date of issuance and shall be implemented
within 45 days of issuance.
Amendment Nos.: 280 and 271.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the TSs.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68745). The January 29, 2003, supplemental letter provided
clarifying information that was within the scope of the initial notice
and did not change the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated February 5, 2003.
No significant hazards consideration comments received: No.
Dated in Rockville, Maryland, this 10th day of February, 2003.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 03-3689 Filed 2-13-03; 8:45 am]
BILLING CODE 7590-01-P