[Federal Register Volume 67, Number 131 (Tuesday, July 9, 2002)]
[Notices]
[Pages 45560-45576]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-16956]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from June 14, 2002 through June 27, 2002. The
last biweekly notice was published on June 25, 2002 (67 FR 42814).
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, 11555 Rockville
Pike (first floor), Rockville, Maryland. The filing of requests for a
hearing and petitions for leave to intervene is discussed below.
By July 25, 2002, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR
[[Page 45561]]
2.714, \1\ which is available at the Commission's PDR, located at One
White Flint North, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records will be accessible from the
Agencywide Documents Access and Management 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 subparagraphs (d)(1) and (2),
regarding petitions to intervene and contentions. Those provisions
are extant and still applicable to petitions to intervene. Those
provisions are as follows: ``In all other circumstances, such ruling
body or officer shall, in ruling on--
(1) A petition for leave to intervene or a request for hearing,
consider the following factors, among other things:
(i) The nature of the petitioner's right under the Act to be
made a party to the proceeding.
(ii) The nature and extent of the petitioner's property,
financial, or other interest in the proceeding.
(iii) The possible effect of any order that may be entered in
the proceeding on the petitioner's interest .
(2) The admissibility of a contention, refuse to admit a
contention if:
(i) The contention and supporting material fail to satisfy the
requirements of paragraph (b)(2) of this section; or
(ii) The contention, if proven, would be of no consequence in
the proceeding because it would not entitle petitioner to relief.''
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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 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, 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, 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].
[[Page 45562]]
Consumers Energy Company, Docket No. 50-155, Big Rock Point Nuclear
Plant, Charlevoix, County, Michigan
Date of amendment request: June 11, 2002.
Description of amendment request: The amendment request changes the
Defueled Technical Specifications by adding applicability statements to
the requirements for storage and inspection of spent fuel and for the
program requirements for spent fuel pool water chemistry.
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 proposed change does not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The requested license amendment involves the addition of
applicability statements to the program and activity requirements
for the storage and inspection of spent fuel activities and
requirements and the SFP [spent fuel pool] water chemistry. These
applicability statements make requirements applicable whenever
irradiated fuel is stored in the SFP. Once irradiated fuel has been
completely removed from the SFP and transferred to a certified dry
fuel storage container under a general 10 CFR Part 72 license, these
program requirements for the SFP are no longer necessary. The
program requirements consist of the specification, establishment,
implementation, and maintenance of fuel configuration, fuel cooling,
and water chemistry for the SFP to minimize the potential effects of
decay heat and corrosion.
The corresponding program requirements for fuel storage in dry
containers are specified in the container's certificate of
conformance and safety analysis report. The corresponding program
requirements currently include:
1. Analysis of fuel assemblies to determine maximum temperatures
within the fuel assemblies to the temperature at the edge of the
assemblies,
2. Design of passive heat removal components to remove heat via
convection, conduction, and radiation, and
3. Specifications for canister vacuum drying pressure and helium
backfill pressure that would ensure that a sufficiently inert
environment is produced within the canister to inhibit corrosion.
The program requirements associated with fuel storage in the SFP
do not contribute to accident prevention or mitigation following the
complete removal of irradiated fuel. The corresponding program
features for fuel storage in dry storage containers are specified
and containers are specified and controlled under other applicable
license documents. These changes do not significantly increase the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any other accident previously evaluated.
The requested amendment involves the addition of applicability
statements that will have the effect of making a program requirement
associated with the SFP inapplicable when the SFP is no longer used
for irradiated fuel storage. The corresponding program requirements
are adequately specified in applicable license documents. The
elimination of this program requirement following complete removal
of irradiated fuel from the SFP does not result in any new or
different accident initiators from those already assumed in
accidents previously evaluated, nor does it exacerbate any such
accidents. Therefore, these changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety.
The safety margins produced as a result of the specification of
program requirements for fuel storage in the SFP are adequately
maintained in corresponding program requirements associated with
fuel storage in dry storage containers. These corresponding program
requirements are specified in the dry storage container's
certificate of compliance and safety analysis report. Therefore,
this change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's significant hazards
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: David A. Mikelonis, Esquire, Consumers
Energy Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Section Chief: Robert A. Gramm.
Dominion Nuclear Connecticut, Inc., et al., Docket Nos. 50-336 and 50-
423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London
County, Connecticut
Date of amendment request: May 13, 2002.
Description of amendment request: The proposed amendment modifies
the Millstone Nuclear Power Station, Unit No. 2 (MP2) and Unit No. 3
(MP3) Technical Specifications (TSs) to change selected MP2 and MP3
radiological-related TSs. These changes are due to the revision to Part
20 of Title 10 of the Code of Federal Regulations.
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 staff's review is
presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
These changes do not have an impact on the acceptance criteria for
any design-basis accident described in the respective MP2 or MP3
Updated Final Safety Analysis Report (UFSAR).
The changes have no impact on plant equipment operation. Since the
changes are administrative or editorial in nature they cannot affect
the likelihood or consequences of accidents. Therefore, the proposed
changes will not increase the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The revisions to the Occupational Radiation Exposure Report,
Radioactive Effluent Controls Program, and High Radiation Area
Specifications in accordance with TSTF travelers 152, 258, and 308 will
have no effect on plant operation. Since the proposed changes are
solely administrative or editorial in nature, they do not affect plant
operation in any way.
The proposed changes do not involve a physical alteration of the
plant or change the plant configuration (no new or different type of
equipment will be installed). The proposed changes do not require any
new or unusual operator actions. The changes do not alter the way any
structure, system, or component functions and do not alter the manner
in which the plant is operated. The changes do not introduce any new
failure modes. Therefore, the proposed changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
Since the proposed changes are solely administrative or editorial
changes to the TSs, they do not affect plant operation in any way. The
proposed changes to each unit's TSs will revise them to reflect the
requirements of the current 10 CFR Part 20, standardize terminology,
provide clearer guidance, clarify inconsistencies, remove extraneous
information, and result in minor format changes that will not result in
any technical changes to current requirements.
The proposed changes have no effect on any safety analyses
assumptions and therefore do not impact any margins of safety. The
proposed changes do not impact any acceptance criteria for the design-
basis accidents described in the
[[Page 45563]]
respective MP2 or MP3 UFSAR and do not impact the consequences of
accidents previously evaluated. Therefore, the proposed changes will
not result in a reduction in a margin of safety.
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: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, CT 06385.
NRC Section Chief: James W. Clifford.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina and Docket
Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2,
Mecklenburg County, North Carolina
Date of amendment request: May 29, 2002.
Description of amendment request: The amendments would revise the
Technical Specifications 5.5.2 to allow, on a one-time basis, extension
of the interval governing the conduct of containment integrated leak
rate test (ILRT) from ten to fifteen years. The amendments represent a
one-time exception to the ten-year frequency of the performance-based
Type A tests as delineated by Regulatory Guide 1.163, ``Performance-
Based Containment Leak-Test Program,'' September 1995. The amendments
will allow conduct of each respective unit's ILRT within fifteen years
from the last ILRT performed for each unit.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The following discussion is a summary of the evaluation of the
changes contained in these proposed amendments against the 10 CFR
50.92(c) requirements to demonstrate that all three standards are
satisfied. A no significant hazards consideration is indicated if
operation of the facility in accordance with the proposed amendments
would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated, or
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated, or
3. Involve a significant reduction in a margin of safety.
First Standard
The proposed amendments will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed extension to the Type A testing intervals
cannot increase the probability of an accident previously evaluated
since extension of the intervals is not a physical plant
modification that could alter the probability of accident
occurrence, nor is it an activity or modification by itself that
could lead to equipment failure or accident initiation. The proposed
extension to the Type A testing intervals does not result in a
significant increase in the consequences of an accident as
documented in NUREG-1493. The NUREG notes that very few potential
containment leakage paths are not identified by Type B and Type C
tests. It concludes that reducing the Type A testing frequency to
once per twenty years leads to an imperceptible increase in risk.
Catawba and McGuire provide a high degree of assurance through
testing and inspection that the containments will not degrade in a
manner detectable only by Type A testing. Recent Type A tests for
the Catawba and McGuire units identified containment leakage within
acceptance criteria, indicating a very leak tight containment.
Inspections required by the ASME Code are also performed in order to
identify indications of containment degradation that could affect
leak tightness. Separately, Type B and Type C testing, required by
TS [Technical Specifications], identify any containment opening from
design penetrations, such as valves, that would otherwise be
detected by a Type A test. These factors establish that an extension
to the Type A test intervals will not represent a significant
increase in the consequences of an accident.
Second Standard
The proposed amendments will not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed revisions to the Catawba and McGuire TS add
a one-time extension to the current interval for Type A testing. The
current test interval of ten years, based on past performance, would
be extended on a one-time basis to fifteen years from the last Type
A test. The proposed extension to Type A test intervals does not
create the possibility of a new or different type of accident since
there are no physical changes being made to the plants and there are
no changes to the operation of the plants that could introduce a new
failure mode.
Third Standard
The proposed amendments will not involve a significant reduction
in a margin of safety. The proposed revisions to the Catawba and
McGuire TS add a one-time extension to the current interval for Type
A testing. The current test interval of ten years, based on past
performance, would be extended on a one-time basis to fifteen years
from the last Type A test. The proposed extension to Type A test
intervals will not significantly reduce the margin of safety. The
NUREG-1493 generic study of the effects of extending containment
leakage testing intervals found that a twenty-year interval resulted
in an imperceptable increase in risk to the public. NUREG-1493 found
that, generically, the design containment leakage rate contributes
about 0.1 percent of the overall risk and that decreasing the Type A
testing frequency would have a minimal effect on this risk, since 95
percent of the Type A detectable leakage paths would already be
detected by Type B and Type C testing. Similar proposed changes have
been previously reviewed and approved by the NRC, and they are
applicable to Catawba and McGuire.
Based upon the preceding discussion, Duke Energy Corporation has
concluded that 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 three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn , Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
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.
Description of amendment request: The proposed change will revise
Appendix 3B and Section 6.2.1.2 of the Updated Safety Analysis Report
pertaining to the method of analysis. The proposed change will replace
the current vendor THREED code for room pressure-temperature analyses
due to High Energy Line Breaks (HELB) with GOTHIC (Generation of
Thermal-Hydraulic Information for Containments). The proposed change
will allow Entergy Operations, Inc. (EOI) to update the analysis and to
evaluate additional changes to the plant.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will the operation of the facility in accordance with these
proposed changes involve a significant increase in the probability
or consequence of an accident previously evaluated?
Response: The proposed change involves no increase in the
probability of the accidents previously evaluated since no physical
change to the plant will be made. The change of the High Energy Line
Break (HELB) analysis method does not affect the probability of the
analyzed event occurring.
[[Page 45564]]
The line break locations have not been affected and remain as
originally designed.
This submittal is required due to the change of HELB analysis
code from the vendor code THREED to the modern industry standard
analysis code GOTHIC. This is a change in the methodology for
determining the effects of the mass and energy release in the plant
as a result of currently postulated events. The change in the
evaluation methodology has been benchmarked and reviewed to confirm
the results remain consistent with the current analysis. The changes
to the model used for the additional analysis allow the use of new,
more physically realistic models for Containment and Auxiliary
Building pressure/temperature responses and will demonstrate
continued qualification of the equipment in these buildings. Mass
and energy releases for some cases have also been recalculated to
credit pipe friction, which was only credited for certain cases
previously.
With these new results the equipment has been reviewed and
remains qualified per current programs established at RBS [River
Bend Station]. Therefore, the plant will continue to function as
designed and thus there will be no impact on consequences.
2. Will the operation of the facility in accordance with these
proposed changes create the possibility of a new or different kind
of accident from any accident previously evaluated?
Response: No physical change to the plant will be made. The HELB
locations were identified by reviewing all the possible break
locations in each Auxiliary and Containment Building volume
containing high-energy lines. The locations of the breaks remain the
same as the previous HELB analyses. The HELB analyses have been
evaluated for the current plant configuration. The new HELB analysis
has been benchmarked against the previous accepted methods and found
to correlate with the previous analysis. Therefore the results can
be used to predict plant responses to events. The proposed change
uses improved methods for mass and energy release calculation and
pressure / temperature responses to determine the EQ [equipment
qualification] qualification envelopes. Therefore, no new or
different interaction would be created.
3. Will the operation of the facility in accordance with these
proposed changes involve a significant reduction in a margin of
safety?
Response: The operation of the facility in accordance with the
proposed changes will not involve a significant reduction in a
margin of safety.
The GOTHIC code has been successfully benchmarked versus the
vendor THREED code, which was used in the original design
calculations. The HELB analysis results with the benchmarking GOTHIC
model are consistent with the THREED results. Therefore, the use of
GOTHIC code will not involve a reduction in an identified margin of
safety. Given that GOTHIC code is an improved methodology and it has
been extensively qualified against the solved analytical problems
and testing results, the use of GOTHIC code will produce more
accurate pressure/temperature responses for the HELB analyses. The
use of the GOTHIC code has been approved for pressure/temperature
responses analysis at various other plants including Joseph M.
Farley Nuclear Plant, Units 1 and 2, and Waterford [Steam Electric
Station, Unit] 3.
The results with the revised methods will be used to show that
safety equipment meets the EQ requirements. The peak temperatures
and pressures in the HELB GOTHIC benchmark model are within the
existing EDC [environmental design criteria] envelopes. Therefore,
the pressure/temperature responses from the HELB benchmark analyses
have no impact on the equipment qualification.
The methodology in the original design calculations is very
conservative. The mass and energy releases without crediting
friction introduce excessive amount of high-energy fluid into the
break rooms, which is unrealistic. Some HELB calculations have
credited both the frictional flows and the additional zone to
eliminate excessive conservatism in the pressure/temperature
responses. There is no reduction in a margin of safety and the
design room differential pressure limits continue to be [met].
The use of this method by EOI RBS is consistent with the
guidance given in NRC [U.S. Nuclear Regulatory Commission] Generic
Letter 83-11 and Supplement 1, addressing the performance of safety
analyses by licensees. EOI has implemented this guidance for the
GOTHIC methodology consistent with the intended application. The
GOTHIC methodology has been verified and validated by the software
vendor. In addition, this methodology is controlled by EOI
procedures and under the EOI quality assurance program. This
includes EOI and RBS specific verification and validation of this
application of GOTHIC and review of the calculations performed.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.
Entergy Nuclear Operations, Inc. (ENO), Docket No. 50-003, Indian Point
Nuclear Generating Station, Unit 1, Buchanan, New York
Date of application for amendment: May 30, 2002.
Description of amendment request: The proposed changes will modify
the Indian Point Generating Station, Unit 1 (IP1), Technical
Specifications (TSs) and Provisional Operating License No. DPR-5. IP1
is completely enclosed within the protected area for Indian Point
Nuclear Generating Station, Unit 2 (IP2). IP1 depends on the IP2 TSs
and processes for the implementation of certain regulatory
requirements. The requested changes will simplify the IP1 TSs to
facilitate the IP2 transition to the Improved TSs. The IP1 TSs will be
reformatted, reordered and repaginated for consistency and clarity. ENO
also proposes that certain changes supersede requirements of the
``Order Approving Decommissioning Plan and Authorizing Decommissioning
of Facility'' \2\ (the Order) to ensure compliance with the current
requirements of 10 CFR Part 50.59, ``Changes, tests, and experiments.''
and 10 CFR Part 50.82, ``Termination of license,'' for evaluating
whether changes can be made to IP1 without NRC approval.
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\2\ U.S. Nuclear Regulatory Commission (NRC) letter to
Consolidated Edison, ``Order to Authorize Decommissioning and
Amendment No. 45 to License No. DPR-5 for Indian Point Unit 1 (TAC
No. M59664),'' dated January 31, 1996.
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Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed license amendment involve a significant
increase in the probability or in the consequences of an accident
previously evaluated?
The NSB [Nuclear Services Building] sewage effluent line
radiation monitor is not required to function to mitigate any
postulated accident. The design or operation of the radiation
monitor on the existing sewage effluent discharge line will not be
changed by deleting operability and surveillance requirements for
the NSB sewage effluent radiation monitor from the IP1 TS. The
nuclear services building sewage effluent line is neither an
accident initiator nor mitigator.
The other proposed changes do not result in a change to the
design or operation of any plant structure, system or component.
Therefore any assumptions of the operability or performance of any
structure, system or component in accident evaluations are
unchanged.
The proposed fire protection TS 2.11 involves deleting
requirements from the IP1 TS that are solely applicable to IP2. Any
assumptions of the operability or performance of any structure,
system or component in IP2 accident evaluations, including the Fire
Plan, are unchanged. Therefore, there is no increase in the
probability or in the consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
The proposed TS change involves the deletion of operability and
surveillance requirements for radioactive effluent monitoring of the
NSB sewage effluent from the IP1 TS. The proposed TS changes do not
[[Page 45565]]
affect the design or operation of any plant structure, system, or
component.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
This change to TS 1.0 does not affect a design function for or
the operation of any plant structure, system, or component. The
change does not affect the method of ENO's compliance with any
regulation.
The proposed TS change involving IP1 TS 2.11 statement governs
the protection of IP2 safe shutdown systems from fire. Effective
protection of IP2 safe shutdown systems from fire is mandated by IP2
License Condition 2.K. The effectiveness of ENO compliance with IP2
License Condition 2.K is not affected by this change. In addition,
this change does not affect a design function or the operation of
any plant structure, system, or component.
The proposed changes to TS sections 3.1 and 3.2 involve
eliminating the duplication of requirements in the IP1 TS and
incorporating the requirements by reference to the IP2 TS. A single
ENO organization operates both IP1 and IP2. The effective
organizational requirements to ensure compliance with all ENO IP1
and IP2 site requirements are mandated by the IP2 TS. The
effectiveness of ENO's safety management of the Indian Point site is
not affected by this change. In addition, this change does not
affect a design function or the operation of any plant structure,
system, or component.
The proposed TS change to sections 4.1 and 5.2 involves
eliminating the reference in the IP1 TS to the specific applicable
section number of the IP2 TS. A single organization operates both
IP1 and IP2. The applicable IP2 TS is obvious by the activity title.
The effectiveness of ENO's safety management of the Indian Point
site is not affected by this change. In addition, this change does
not affect a design function or the operation of any plant
structure, system, or component.
Effective compliance with the 10CFR20 requirements for radiation
protection and monitoring radioactive effluent releases is mandated
by other IP1 and IP2 TS and license provisions. The effectiveness of
ENO compliance with 10CFR20 requirements is not adversely affected
by the elimination of TS requirements for the radiation protection
plan and radioactive effluent monitoring on the nuclear services
building sewage effluent line.
The proposed TS change involves requirements for the site
Meteorological Monitoring and Radiological Environmental Monitoring
programs. However, IP2 TS provisions mandate effective compliance
for meteorological and radiological environmental monitoring. The
effectiveness of ENO compliance with 10CFR50.47, 10CFR100, and
10CFR20 requirements is not adversely affected by this change. In
addition, this change does not affect a design function or the
operation of any plant structure, system, or component. IP2 TS
provisions mandate effective compliance with requirements for
radiation protection. The effectiveness of ENO's compliance with 10
CFR 20 is not adversely affected by this change or the change to the
section for sealed sources. In addition, this change does not affect
a design function or the operation of any plant structure, system,
or component.
The proposed TS change involves the location of routine and
event reporting requirements. However, other IP2 TS provisions
mandate effective compliance with reporting requirements. In
addition, this change does not affect a design function or the
operation of any plant structure, system, or component.
The effectiveness of ENO's compliance with 10CFR50.59 is not
adversely affected by the clarification and relocation of the
applicability of the FSAR [Final Safety Analysis Report]. In
addition, this change does not affect a design function or the
operation of any plant structure, system, or component.
Therefore, the change does not result in a change to any of the
safety analyses or any margin of safety.
ENO also requests that the expiration date of IP1 Provisional Operating
License No. DPR-5 be changed from ``midnight, October 14, 2002,'' to
``midnight, September 28, 2013,'' the current expiration date for
Facility Operating License No. DPR-26 for IP2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed license amendment involve a significant
increase in the probability or in the consequences of an accident
previously evaluated?
In its Safety Evaluation and Environmental Assessment for its
January 31, 1996, Order Approving Decommissioning Plan and
Authorizing Decommissioning of Facility, the NRC evaluated the
acceptability of the possession-only license and safety issues
related to SAFSTOR of Indian Point Nuclear Generating Unit No. 1
until September 28, 2013. The requested change does not involve any
activity that could change the assumptions of the prior Safety
Evaluation and Environmental Assessment.
Therefore, the proposed license amendment does not involve a
significant increase in the probability or in the consequences of an
accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
In its Safety Evaluation and Environmental Assessment for its
January 31, 1996, Order Approving Decommissioning Plan and
Authorizing Decommissioning of Facility, the NRC evaluated the
acceptability of the possession-only license and safety issues
related to SAFSTOR of Indian Point Nuclear Generating Unit No. 1
until September 28, 2013. The requested change does not involve any
activity that could change the assumptions of the prior Safety
Evaluation and Environmental Assessment.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
In its Safety Evaluation and Environmental Assessment for its
January 31, 1996, Order Approving Decommissioning Plan and
Authorizing Decommissioning of Facility, the NRC evaluated the
acceptability of the possession-only license and safety issues
related to SAFSTOR of Indian Point Nuclear Generating Unit No. 1
until September 28, 2013. The requested change does not involve any
activity that could change the assumptions of the prior Safety
Evaluation and Environmental Assessment.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analyses and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for Licensee: Mr. John Fulton, Assistant General Consul,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Robert A. Gramm.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: May 30, 2002.
Description of amendment request: The proposed amendment would
revise the Facility Operating License and Technical Specifications
(TSs) to increase the licensed core thermal power level to 3067.4
megawatts (MWt), which is a 1.4% increase above the currently
authorized power level of 3025 MWt. The proposed power uprate involves
the improvement in the core power uncertainty allowance originally
required for the emergency core cooling system (ECCS) evaluations
performed in accordance with Appendix K, ``ECCS Evaluation Models,'' to
Part 50 of Title 10 of the Code of Federal Regulations. In addition,
changes would be made in TS Sections 2.2, 3.3, 3.4, 3.7, and the
applicable TS Bases would be revised to account for the change in power
level.
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 45566]]
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The evaluations associated with this proposed change to core
power level have demonstrated that all applicable acceptance
criteria for plant systems, components, and analyses (including the
Final Safety Analysis Report Chapter 14 safety analyses) will
continue to be met for the proposed 1.4% increase in licensed core
thermal power for IP3 [Indian Point Unit 3]. The subject increase in
core thermal power will not result in conditions that could
adversely affect the integrity (material, design, and construction
standards) or the operational performance of any potentially
affected system, component or analysis. Therefore, the probability
of an accident previously evaluated is not affected by this change.
The subject increase in core thermal power will not adversely affect
the ability of any safety-related system to meet its intended safety
function. Further, the radiological dose evaluations in support of
this power uprate effort show that the current FSAR [Final Safety
Analysis Report] Chapter 14 radiological analyses are unaffected,
and that the current dose analyses of record bound plant operation
with the subject increase in licensed core thermal power level.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The evaluations of this proposed amendment show that all
applicable acceptance criteria for plant systems, components, and
analyses (including FSAR Chapter 14 safety analyses) will continue
to be met for the proposed 1.4% power increase in IP3 licensed core
thermal power. The subject increase in core thermal power will not
result in conditions that could adversely affect the integrity
(material, design, and construction standards) or operational
performance of any potentially affected system, component, or
analyses. The subject increase in core thermal power will not
adversely affect the ability of any safety-related system to meet
its safety function. Furthermore, the conditions associated with the
subject increase in core thermal power will neither cause initiation
of any accident, nor create any new credible limiting single
failure. The power uprate does not result in changing the status of
events previously deemed to be non-credible being made credible.
Additionally, no new operating modes are proposed for the plant as a
result of this requested change.
Therefore, the subject increase in core thermal power level will
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The evaluations associated with this proposed change show that
all applicable acceptance criteria for plant systems, components,
and analyses (including FSAR Chapter 14 safety analyses) will
continue to be met for this proposed 1.4% increase in IP3 licensed
core thermal power. The subject increase in core thermal power will
not result in conditions that could adversely affect the integrity
(material, design, and construction standards) or operational
performance of any potentially affected system, component, or
analysis. The subject power uprate will not adversely affect the
ability of any safety-related system to meet its intended safety
function. For example, most IP3 analyses already add a 2%
uncertainty allowance to the nominal power level to account solely
for power measurement uncertainty. These analyses have not been
revised for the 1.4% uprate power level conditions because the sum
of increased core power level (1.4%) and the improved power
measurement accuracy (uncertainty less than 0.6%) is already bounded
by the currently analyzed 2% uncertainty allowance.
Therefore, the subject increase in core thermal power will not
involve a reduction in [a] margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: June 3, 2002.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.4.9, ``Pressurizer,'' to increase
the pressurizer water level limit when the plant is in Mode 3 (Hot
Standby). The current pressurizer water level limit is applicable for
Modes 1, 2, and 3, and will remain unchanged for Modes 1 and 2. The
proposed amendment would also revise TS 3.8.4, ``DC Sources--
Operating,'' to remove the notes that refer to the one-time amendment
allowing the online replacement of station batteries 31 and 32. The
notes are no longer applicable since the batteries have been replaced.
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.
Pressurizer water level is an assumed initial condition for
certain accident analyses. Plant initial conditions are not accident
initiators and do not have an effect on the probability of the
accident occurring. The proposed change only revises the specified
limit on water level in the pressurizer, so that this change would
not affect accident probability.
The specific accidents for which pressurizer water level is an
assumed initial condition are a loss of load and a loss of normal
feedwater. The limiting accident analysis results occur at full
power conditions when the available core thermal power is maximized.
The proposed change does not affect the specified pressurizer level
limit at any power level from zero to full power. That is, the
pressurizer level limit is not being changed in Modes 1 and 2. The
proposed change does revise the specified pressurizer water level
limit in Mode 3 (Hot Standby) but this does not affect accident
analysis results because the limiting analyses will remain those
that are postulated to occur in Mode 1 with the plant at full power.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve physical changes to
existing plant equipment or the installation of any new equipment.
The design of the pressurizer, the pressurizer level control system
and the pressurizer safety valves is not being changed and the
ability of these systems, structures, and components to perform
their design or safety functions is not being affected. The proposed
change revises the specified limit on pressurizer water level in
Mode 3 (Hot Standby) to allow operators greater flexibility in
performing a plant cooldown. The method used in performing the plant
cooldown is not being changed. This proposed change does not create
new failure modes or malfunctions of plant equipment nor is there a
new credible failure mechanism.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Pressurizer level is an initial condition assumed in certain
accident analyses involving an insurge in the pressurizer and an
increasing reactor coolant system (RCS) pressure. These analyses
demonstrate that the design pressure for the RCS is not exceeded for
the limiting analyses based on the plant at full power. The proposed
change does not affect the existing Technical
[[Page 45567]]
Specification requirement for Mode 1 (Power Operation) or Mode 2
(Plant Startup) and therefore does not affect the assumptions or
results of these accident analyses. The margin for RCS design
pressure demonstrated by these analysis results is not being
reduced. The proposed change only applies to the pressurizer level
limit in Mode 3 (Hot Standby) when there is substantially lower
thermal energy available to cause rapid expansion of reactor coolant
and an insurge to the pressurizer. Protection of the RCS pressure
boundary is still maintained by the pressurizer safety valves, which
are not being modified by the proposed change in pressurizer water
level.
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: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: June 5, 2002.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to implement the alternate
source term methodology for the fuel-handling accident analysis.
Specifically, the proposed amendment would revise TS 3.9.3,
``Containment Penetrations,'' to: (1) Permit the equipment hatch
opening and the personnel air lock doors to be capable of being closed
during movement of irradiated fuel, (2) allow use of administrative
controls for unisolating containment penetrations during movement of
irradiated fuel, (3) delete the containment purge and containment
pressure relief requirements and associated surveillances with the
reactor subcritical for less than 550 hours, and (4) eliminate the TS
applicability ``during core alterations.'' In this regard, the proposed
amendment would adopt TS Task Force (TSTF) Standard TS Change Travelers
TSTF-68, ``Containment Personnel Airlock Doors Open During Fuel
Movement,'' TSTF-312, ``Administratively Control Containment
Penetrations,'' and, in part, TSTF-51, ``Revise Containment
Requirements During Handling Irradiated Fuel and Core Alterations.''
The proposed amendment would also relocate the requirements in TS
3.7.13, ``Fuel Storage Building Emergency Ventilation System,'' and TS
3.3.8, ``Fuel Storage Building Emergency Ventilation System Actuation
Instrumentation,'' to the licensee-controlled Technical Requirements
Manual.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change involves the reanalysis of a fuel handing
accident (FHA) in containment and in the fuel storage building. The
new analysis, based on the Alternate Source Term (AST) in accordance
with 10 CFR [Code of Federal Regulations] 50.67, will replace the
existing analysis based on methodologies and acceptance criteria in
place when Indian Point 3 was originally licensed. As a result of
the new analysis, changes to the Technical Specifications are
proposed which take credit for the new analysis results.
The proposed changes to the technical specifications modify
requirements regarding containment closure during movement of
irradiated fuel assemblies in containment and relocate requirements
for the fuel storage building emergency ventilation system from the
technical specifications to a licensee controlled document. The
proposed changes do not involve physical modifications to plant
equipment and do not change the operational methods or procedures
used for moving irradiated fuel assemblies. As such, there are no
accident initiators affected by the proposed amendment. The revised
requirements apply only when the plant is in a refueling condition
(Mode 6), and specifically only when irradiated fuel is being moved.
Previously evaluated accidents with the plant in other conditions
ranging from cold shutdown (Mode 5) through power operation (Mode 1)
are not affected. The AST methodology is used to evaluate a[n] FHA
that is postulated to occur during fuel movement activities in the
containment building and the fuel storage building. The analysis
follows the guidance of the NRC Regulatory Guide 1.183 and uses the
acceptance criteria of the NRC Standard Review Plan (NUREG 0800) for
offsite doses and General Design Criteria 19 for control room
personnel. The analysis demonstrates that the dose consequences meet
regulatory acceptance criteria. The accident analysis conservatively
assumes that the containment building and the fuel storage building,
including ventilation filtration systems for those building[s] does
not diminish or delay the assumed fission product release. The
analysis does take credit for, and technical specifications enforce,
the presence of 23 feet of water over the irradiated fuel while fuel
movement activities are being performed. The analysis also takes
credit for, and the technical specification bases enforce a fuel
decay time of at least 84 hours. In addition, administrative
controls are put in place to provide for closure of containment
openings in the event of a[n] FHA. Use of an alternate analysis
method does not affect fuel parameters or the equipment used to
handle the fuel. The proposed changes to the technical
specifications reflect assumptions made in the analysis.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment involves the use of an alternate analysis
methodology for the evaluation of the dose consequences from a[n]
FHA that is postulated to occur in either the containment building
or the fuel storage building (FSB). The analysis demonstrates that
containment closure conditions and operation of the containment
purge filtration system are not required to maintain dose
consequence within regulatory limits following a postulated FHA in
containment. Therefore the new analysis supports proposed changes to
requirements for containment closure during movement of irradiated
fuel assemblies in containment. The analysis results also
demonstrate that operation of the fuel storage building emergency
ventilation system is not required to maintain dose consequences
within regulatory limits following a postulated FHA in the FSB. The
containment closure components (e.g., equipment hatch, personnel
airlock doors, and various containment penetrations) and filtration
systems are not accident initiators. The proposed changes do not
involve the addition of new systems or components nor do they
involve the modification of existing plant systems. The proposed
changes do not affect the way in which a[n] FHA is postulated to
occur.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The existing dose analysis methodology and assumptions
demonstrates that the dose consequences of a[n] FHA are within
regulatory limits for whole body and thyroid doses as established in
10 CFR 100. The alternate dose analysis methodology and assumptions
also demonstrates that the dose consequences of a[n] FHA are within
regulatory limits. The limits applicable to the alternate analysis
are established in 10 CFR 50.67 in conjunction with the TEDE (total
effective dose equivalent) acceptance directed in Regulatory Guide
1.183. The acceptance criteria for both dose analysis methods have
been developed for the
[[Page 45568]]
purpose of evaluating design basis accidents to demonstrate adequate
protection of public health and safety. An acceptable margin of
safety is inherent in both types of acceptance criteria.
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: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: June 7, 2002.
Description of amendment request: The proposed amendment would
change the requirements associated with handling irradiated fuel and
performing core alterations. Specifically, the changes would eliminate
operability requirements for secondary containment when handling
recently irradiated fuel and during core alterations. The amendment
would also revise the requirements associated with equipment whose
performance is not credited in the new calculations.
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 not involve a significant increase in the probability or
consequences of an accident previously analyzed?
Response: No.
The proposed TS [Technical Specifications] changes do not modify
the design or operation of equipment used to move spent fuel or to
perform core alterations. Because the equipment affected by the
change is not an initiator to any previously analyzed accident, the
proposed change cannot increase the probability of any previously
analyzed accident.
The conservative re-analysis of the fuel handling accident
concludes that radiological consequences are within the acceptance
criteria in Regulatory Guide 1.183 and 10 CFR 50.67. The results of
the core alteration events, other than the fuel handling accident,
remain unchanged from the original design-basis, which showed that
these events do not result in fuel cladding damage or radioactive
release. The radiological analysis uses the same FHA [fuel handling
accident] source activity previously accepted in the design-basis
FHA analysis. The same source activity is used with the guidance in
the Regulatory Guide 1.183, Appendix B and the passive release/
transport path, which does not take the dose mitigation credit of
engineered safeguards including secondary containment and CREVAS
[Control Room Emergency Ventilation] Systems.
Therefore, this proposed amendment does not involve a
significant increase in the probability of occurrence or
consequences of an accident previously analyzed.
2. Does not create the possibility of a new or different kind of
accident from any accident previously analyzed?
Response: No.
The proposed post-FHA activity transport path is passive in
nature and it does not take the credit of dose mitigation functions
previously credited in the design-basis FHA analysis. The proposed
changes do not introduce any new modes of plant operation and do not
involve physical modifications to the plant.
Therefore, this proposed amendment does not create the
possibility of a new or different kind of accident from any
previously analyzed.
3. Does not involve a significant reduction in [a] margin of
safety?
Response: No.
The proposed changes revise the FitzPatrick TS to establish
operational conditions where specific activities represent
situations during which significant radioactive releases can be
postulated. These new operational conditions are consistent with the
proposed design-basis accident analysis and are established such
that the radiological consequences are less than the regulatory
allowable limits. Safety margins and analytical conservatisms are
retained to ensure that the analysis adequately bounds all
postulated event scenarios. The selected assumptions and release
models provide an appropriate and prudent safety margin against
unpredicted events in the course of an accident and compensates for
large uncertainties in facility parameters, accident progression,
radioactive material transport and atmospheric dispersion. The
proposed TS applicability statements continue to ensure that the
TEDE [Total Effective Dose Equivalent] at the control room and the
exclusion area and low population zone boundaries are below the
corresponding regulatory allowable limits in 10 CFR 50.67(b)(2).
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: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Section Chief: Richard J. Laufer.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Dockets Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
County, Pennsylvania
Date of application for amendments: May 24, 2002.
Description of amendment request: Exelon Generation Company, LLC,
the licensee, is proposing changes to the Peach Bottom Atomic Power
Station, Units 2 and 3 (PBAPS), Operating Licenses and Technical
Specifications associated with an increase in the licensed power level.
The changes involve a proposed 1.62 percent increase in the licensed
reactor core thermal power level (an increase in reactor power level
from 3,458 megawatts thermal to 3,514 megawatts thermal). These changes
result from increased accuracy of the feedwater flow and temperature
measurements to be achieved by utilizing high accuracy ultrasonic flow
measurement instrumentation. This results in a more accurate
determination of reactor core thermal power level. The basis for this
change is consistent with the revision, issued in June 2000, to
Appendix K to Part 50 of Title 10 of the Code of Federal Regulations,
allowing operating reactor licensees to use an uncertainty factor of
less than 2 percent of rated reactor thermal power in analyses of
postulated design-basis loss-of-coolant accidents.
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 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. Evaluation of accident analyses confirmed the effects
of the proposed uprate are bounded by the current dose analyses. All
systems will function as designed, and all performance requirements
for these systems have been evaluated and found acceptable.
The primary loop components (reactor vessel, reactor internals,
control rod drive housings, piping and supports, recirculation
pumps, etc.) continue to comply with their applicable structural
limits and will continue to perform their intended design functions.
Thus, there is no increase in the probability of a structural
failure of these components.
All of 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
continue to
[[Page 45569]]
meet their applicable structural limits and will continue to perform
their intended design functions. Thus, there is no increase in the
probability of a structural failure of these components. All of the
NSSS/BOP interface systems will continue to perform their intended
design functions. The safety relief valves and containment isolation
valves 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, it is 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 3514 megawatts thermal (MWt). Other
analyses performed at a nominal power level have either been
evaluated or re-performed for the 1.62% increased power level. The
results demonstrate that the applicable analysis acceptance criteria
continue to be met at the 1.62% uprate conditions. As such, all
PBAPS Updated Final Safety Analysis Report (UFSAR) Chapter 14
accident analyses continue to demonstrate compliance with the
relevant event acceptance criteria. Those analyses performed to
assess the effects of mass and energy releases remain valid. The
source terms used to assess radiological consequences have been
reviewed and determined to bound operation at the 1.62% uprated
condition.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No. No new accident scenarios, failure mechanisms, or
limiting single failures are introduced as a result of the proposed
changes. All systems, structures, and components previously required
for the mitigation of a transient remain capable of fulfilling their
intended design functions. The proposed changes have no adverse
effects on any safety-related system or component and do not
challenge the performance or integrity of any safety related system.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. Operation at the uprated power condition does not
involve a significant reduction in a margin of safety. Analyses of
the primary fission product barriers have concluded that all
relevant design criteria remain satisfied, both from the standpoint
of the integrity of the primary fission product barrier and from the
standpoint of compliance with the required acceptance criteria. As
appropriate, all evaluations have been performed using methods that
have either been reviewed and approved by the NRC, or that 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: Mr. Edward Cullen, Vice President and
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way,
Kennett Square, PA 19348.
NRC Section Chief: James W. Clifford.
Exelon Generation Company, LLC, Docket No. 50-254, Quad Cities Nuclear
Power Station, Unit 1, Rock Island County, Illinois
Date of amendment request: May 30, 2002.
Description of amendment request: The proposed change revises the
safety limit minimum critical power ratio for Unit 1 Cycle 18 for two
loop operation and single loop operation.
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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits have been established consistent with NRC
approved methods to ensure that fuel performance during normal,
transient, and accident conditions is acceptable. The proposed
change conservatively establishes the safety limit for the minimum
critical power ratio (SLMCPR) for Quad Cities Nuclear Power Station
(QCNPS), Unit 1, Cycle 18 such that the fuel is protected during
normal operation and during any plant transients or anticipated
operational occurrences.
Changing the SLMCPR does not increase the probability of an
evaluated accident. The change does not require any physical plant
modifications, physically affect any plant components, or entail
changes in plant operation. Therefore, no individual precursors of
an accident are affected.
The proposed change revises the SLMCPR to protect the fuel
during normal operation as well as during any transients or
anticipated operational occurrences. Operational limits will be
established based on the proposed SLMCPR to ensure that the SLMCPR
is not violated during all modes of operation. This will ensure that
the fuel design safety criteria (i.e., that at least 99.9% of the
fuel rods do not experience transition boiling during normal
operation and anticipated operational occurrences) is met. Since the
operability of plant systems designed to mitigate any consequences
of accidents has not changed, the consequences of an accident
previously evaluated are not expected to increase.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Creation of the possibility of a new or different kind of accident
would require the creation of one or more new precursors of that
accident. New accident precursors may be created by modifications of
the plant configuration, including changes in allowable modes of
operation. The proposed change does not involve any modifications of
the plant configuration or allowable modes of operation. The
proposed change to the SLMCPR assures that safety criteria are
maintained for QCNPS, Unit 1, Cycle 18.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
Does the proposed change involve a significant reduction in a
margin of safety?
The value of the proposed SLMCPR provides a margin of safety by
ensuring that no more than 0.1% of the rods are expected to be in
boiling transition if the MCPR limit is not violated. The proposed
change will ensure the appropriate level of fuel protection.
Additionally, operational limits will be established based on the
proposed SLMCPR to ensure that the SLMCPR is not violated during all
modes of operation.
This will ensure that the fuel design safety criteria (i.e.,
that at least 99.9% of the fuel rods do not experience transition
boiling during normal operation as well as anticipated operational
occurrences) are met.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
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.
[[Page 45570]]
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: June 13, 2002.
Description of amendment request: The amendment would revise the
Improved Technical Specifications (ITS) 3.3.8 and associated bases,
``Emergency Diesel Generator (EDG) Loss of Power Start (LOPS),'' by
changing the completion time for required action D.2 from 12 to 36
hours. The amendment also corrects a typographical error in ITS 3.3.8
and clarifies the discussion in Bases Section B 3.3.8 for Actions D.1
and D.2 to recognize the applicability of ITS 3.3.8 in MODES 5 and 6.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does not involve a significant increase in the probability
or consequences of an accident previously analyzed.
The proposed license amendment revises the Required Time to
place the plant in MODE 5 if an inoperable loss of voltage Function
for the emergency diesel generator (EDG) loss of power start (LOPS)
cannot be restored to OPERABLE status, corrects a typographical
error in the Section Number of ITS 3.3.8, and clarifies the wording
of ITS Bases Section B 3.3.8 for Action D.1 and D.2 regarding the
applicability of the specification during MODES 5 and 6.
The EDG LOPS is intended to protect engineered safeguards
equipment from damage due to sustained undervoltage conditions, and
to ensure rapid restoration of power to the engineered safeguards
electrical buses in the event of a loss of offsite power. The EDG
LOPS is not an initiator of any design basis accident. The design
functions of the EDG LOPS and the initial conditions for accidents
that require an EDG LOPS will not be affected by the change.
Therefore, the change will not increase the probability or
consequences of an accident previously evaluated.
(2) Does not create the possibility of a new or different kind
of accident from any accident previously analyzed.
The proposed amendment involves no changes to the design
functions or operation of the EDG LOPS. Editorial corrections,
clarification of the wording in Bases Section B 3.3.8, or changing
the Required Completion Time for placing the plant in MODE 5 when an
inoperable loss of voltage function cannot be restored will not
introduce any new failure mechanisms, malfunctions or accident
initiators. Therefore, the proposed change will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) Does not involve a significant reduction in the margin of
safety.
The proposed change corrects a typographical error, clarifies
the wording of Bases Section B 3.3.8 for Actions D.1 and D.2, and
revises the required Completion Time to place the plant in MODE 5.
The revised Completion Time will allow the plant to be shutdown in
an orderly fashion without challenging plant systems or plant
cooldown limits. The proposed change does not change the design or
operation of the EDG LOPS, and does not impact the ability of the
EDG LOPS to perform its design functions. Thus, the proposed
amendment will not result in a reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: R. Alexander Glenn, Associate General
Counsel (MAC-BT15A), Florida Power Corporation, P.O. Box 14042, St.
Petersburg, Florida 33733-4042.
NRC Acting Section Chief: Kahtan N. Jabbour.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: June 7, 2002.
Description of amendment request: The proposed amendments would
delete requirements from the Technical Specifications (TSs) (and, as
applicable, other elements of the licensing bases) to maintain a Post
Accident Sampling System (PASS). Licensees were generally required to
implement PASS upgrades as described in NUREG-0737, ``Clarification of
TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power
Plants to Assess Plant and Environs Conditions During and Following an
Accident.'' Implementation of these upgrades was an outcome of the
lessons learned from the accident that occurred at TMI, Unit 2.
Requirements related to PASS were imposed by Order for many facilities
and were added to or included in the TSs for nuclear power reactors
currently licensed to operate. However, lessons learned and
improvements implemented over the last 20 years have shown that the
information obtained from PASS can be readily obtained through other
means, or is of little use in the assessment and mitigation of accident
conditions.
The Nuclear Regulatory Commission (NRC) staff issued a notice of
opportunity for comment in the Federal Register on December 27, 2001
(66 FR 66949) on possible amendments to eliminate PASS, including a
model safety evaluation and model no significant hazards consideration
(NSHC) determination, using the consolidated line item improvement
process. The NRC staff subsequently issued a notice of availability of
the models for referencing in license amendment applications in the
Federal Register on March 20, 2002 (67 FR 13027). The licensee affirmed
the applicability of the NSHC determination in its application dated
June 7, 2002. The NSHC determination is restated below.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of NSHC is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
[[Page 45571]]
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radioisotopes
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in [a] Margin of Safety
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in [a] margin of safety.
Based upon the reasoning presented above and the previous discussion of
the amendment request, the requested change does not involve a
significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Richard J. Laufer.
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: May 29, 2002.
Description of amendment request: The proposed amendment would
revise TS 3.8.1, ``AC Sources--Operating,'' to allow portions of
Surveillance Requirement (SR) 3.8.1.5 to be performed with the units in
Mode 1, 2, 3 or 4. This proposed amendment is consistent with changes
made to NUREG-1431, Standard Technical Specifications, Westinghouse
Plants, by Technical Specification Task Force (TSTF) Traveler, TSTF-
283, Revision 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 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 standby emergency power sources are primarily a support
system for systems required to be operable for accident mitigation.
SR 3.8.1.5 demonstrates the standby emergency power source
operation, during a loss of offsite power actuation test signal in
conjunction with an Engineering Safeguards Feature (ESF) actuation
signal. The proposed amendment only changes the allowed operating
Modes in which portions of this surveillance may be performed.
Performing portions of the surveillance in Mode 1, 2, 3, or 4 will
require an assessment to determine that plant safety is maintained
or will be enhanced.
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. These changes do not introduce any new or different
normal operation or accident initiators. Performing the surveillance
in Mode 1, 2, 3, or 4 will require an assessment to determine that
plant safety is maintained or will be enhanced.
Equipment important to safety will continue to operate as
designed. 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. Therefore, operation of the Point Beach Nuclear
Plant in accordance with the proposed amendment 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.
The standby emergency power sources are primarily a support
system for systems required to be operable for accident mitigation.
SR 3.8.1.5 demonstrates the standby emergency power source
operation, during a loss of offsite power actuation test signal in
conjunction with an ESF actuation signal. Performing the
surveillance in Mode 1, 2, 3, or 4 will require an assessment to
determine that plant safety is maintained or will be enhanced. 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, allowing a portion
of the surveillance to be performed in Mode 1, 2, 3, or 4, 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.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: May 22, 2002.
Description of amendment request: The proposed amendment revises
Technical Specifications (TSs) 3/4.3.5, allowing the automatic
operation of the atmospheric steam relief valves during Mode 2 to
maintain secondary side pressure at or below an indicated steam
generator pressure of 1225 psig during startup and shutdown of the
reactors.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change only provides another method of controlling
the SG PORVs [steam generator power-operated relief valves] under
specified operating conditions. The operating conditions in
Specification 3/4.3.5 remain unchanged. No change is required to
plant design since the proposed method of control is already part of
the plant's configuration. The proposed method of control is the
same method of control
[[Page 45572]]
normally required by the specification in Modes 1 and 2. The
proposed method of control will not impact the accident analysis
assumptions or results. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed method of controlling the SG PORVs is the same
method that these valves are controlled in Modes 1 and 2 by the
specification under normal conditions. The proposed change will
allow the setpoint of these valves to be adjusted to support startup
and shutdown activities. The adjustment of the setpoint is
restricted so that the accident analysis is not impacted. No change
to the design of the valves or plant configuration is required to
implement the proposed change. Therefore, the proposed change does
not create the possibility of a new or different kind of accident
from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change that will allow for an additional method of
controlling the SG PORVs during startup and shutdown activities is
consistent with the operating restrictions for the current method of
valve control. The accident analysis assumptions and results will
remain unaffected. 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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis, &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: Robert A. Gramm.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: May 23, 2002.
Description of amendment request: The proposed amendment revises
the near-end of life (EOL) Moderator Temperature Coefficient (MTC)
Surveillance Requirements by placing a set of conditions on core
operation.
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 probability or consequences of accidents previously
evaluated in the UFSAR [updated final safety analysis report] are
unaffected by this proposed change because there is no change to any
equipment response or accident mitigation scenario. There are no
additional challenges to fission product barrier integrity.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed change.
The proposed change 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. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety associated with the acceptance criteria of
any accident is unchanged. The proposed change will have no affect
on the availability, operability, or performance of the safety-
related systems and components. A change to a surveillance
requirement is proposed, but the limiting conditions for operation
required by the Technical Specifications are not changed.
The Technical Specifications Bases are founded in part on the
ability of the regulatory criteria to be satisfied assuming the
limiting conditions for operation are met for the various systems.
Conformance to the regulatory criteria for operation with the
conditional exemption from the near-EOL MTC measurement is
demonstrated and the regulatory limits are not exceeded. Therefore,
the margin of safety as defined in the TS [technical specification]
is not reduced and 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 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.
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: May 24, 2002.
Description of amendment request: The proposed amendments would
allow Mode 2 (startup) operation with two, rather than three,
intermediate range monitor channels per trip system.
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 intermediate range monitors (IRMs) monitor neutron flux
levels in the reactor core during startup. The IRM detectors are
capable of generating a trip signal during a continuous rod
withdrawal error in the startup range. However, the IRMs perform no
function related to the probability of occurrence of a previously
evaluated accident. Also, the IRM trip signal is not necessary to
mitigate the limiting control rod withdrawal error. The limiting
case assumes the trip signal is generated from the safety-related
average power range monitor (APRM). Therefore, the consequences of
this previously evaluated abnormal operating transient are not
increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change reduces the number of required operable IRM
channels per trip system from three to two. However, the manner in
which the actuation logic functions and the systems respond are
unaffected by the proposed change. Furthermore, the IRMs will
continue to perform their design function of core monitoring during
startup and mitigating nonlimiting transient events postulated to
occur during startup. Therefore, the proposed change cannot 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 Bases for Units 1 and 2 Technical Specification Table
3.3.1.1-1 state the ``IRMs are capable of generating trip signals
that can be used to prevent fuel damage resulting from abnormal
operating transients in the intermediate power (startup) range.''
The proposed change ensures the IRMs will still effectively mitigate
these events. The most significant source of reactivity change is
due to a control rod withdrawal error. With the proposed change, the
IRMs will continue to
[[Page 45573]]
provide protection against rod withdrawal errors, and peak fuel
energy depositions will remain below the 170 cal/gm threshold
criterion defined in the Technical Specifications Bases. Therefore,
the proposed change does not reduce 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.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: March 19, 2002, as supplemented on
June 3, 2002.
Description of amendment request: The proposed Technical
Specification changes involve the removal of the existing scram
function and Group 1 isolation valve closure functions of the Main
Steam Line Radiation Monitors (MSLRM). An explicit requirement for
periodic functional test and calibration of the MSLRM is added to
maintain operability of the mechanical vacuum pump (MVP) isolation
function. This proposed no significant hazards consideration
determination replaces in its entirety the notice published in the
Federal Register on May 14, 2002 (67 FR 34495).
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. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The scram and Group 1 isolation functions of the MSLRMs do not
serve as initiators for any of the accidents evaluated in the
Updated Final Safety Analysis Report (UFSAR). The MSLRM scram
function is not credited in the UFSAR, and the Group 1 isolation
trip function of the MSLRMs was only assumed in one design-basis
event which was the control rod drop accident. Because these
functions are not initiators of accidents, their removal does not
increase the probability of occurrence of previously evaluated
accidents.
There is no accident analysis that relies on the high radiation
scram of the reactor protection system and its removal has no impact
on the consequences of accidents previously evaluated. The results
of the control rod drop accident analysis remain within approved
guidelines, thus any potential increase in consequences would not be
considered significant.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Create the possibility for a new or different kind of
accident from any previously evaluated.
The proposed changes to the plant involve limited changes to
protective circuitry, but do not involve any plant hardware changes
that could introduce any new failure modes. The changes will not
affect non-MSLRM scram and isolation functions. In addition, the
MSLRMs will remain active for other trip/isolation functions, and
these monitors will still alarm in the control room to alert
operators to off-normal conditions.
Therefore, the removal of the Group 1 isolation valve closure
and scram functions of the MSLRMs does not create the possibility of
a new or different kind of accident than those previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change involves the elimination of the scram and
Group I isolation signal from the MSLRMs. Operation under the
proposed change will not change any plant operation parameters, nor
any protective system setpoints other than removal of these
functions. The effects of the control rod drop accident without the
MSLRM scram and isolation signal results in doses which remain well
within 10 CFR Part 100, ``Reactor Site Criteria,'' limits.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Based on this review, it appears that the three standards of 10
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
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.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: June 13, 2002.
Brief description of amendment request: The proposed amendment
would revise Technical Specifications Section 4.13.A, ``Inspection
Requirements,'' to allow the use of the optimum eddy current probe
size when performing steam generator tube inspections. The proposed
amendment would also correct several grammatical errors.
Date of publication of individual notice in Federal Register:
June 25, 2002 (67 FR 42806).
Expiration date of individual notice: July 25, 2002.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission
has determined for each of these amendments that the application
complies with the standards and requirements of the Atomic Energy
Act of 1954, as amended (the Act), and the Commission's rules and
regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with
these actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR
51.22(b), no environmental impact statement or environmental
assessment need be prepared for these amendments. If the Commission
has prepared an environmental assessment under the special
circumstances provision in 10 CFR 51.12(b) and has made a
determination based on that assessment, it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the
Commission's related letter, Safety Evaluation and/or Environmental
Assessment as indicated. All of these items are available for public
inspection at the Commission's Public Document Room, located at One
White Flint North, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records will be accessible from the
Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to
ADAMS or if there are problems in accessing the documents located in
ADAMS, contact the NRC Public Document Room (PDR) Reference staff at
1-800-397-4209, 301-415-4737 or by email to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power
Station, Unit 1, DeWitt County, Illinois
Date of application for amendment: April 17, 2001.
Brief description of amendment: The amendment makes editorial
and administrative corrections to Technical Specifications (TS)
Section 3.3,
[[Page 45574]]
``Instrumentation,'' and eliminates minor discrepancies between TS
Section 3.3 and other plant licensing basis documents.
Date of issuance: June 25, 2002.
Effective date: As of the date of issuance and shall be
implemented within 30 days.
Amendment No.: 152.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 26, 2001
(66 FR 66463). The Commission's related evaluation of the amendment
is contained in a Safety Evaluation dated June 25, 2002.
No significant hazards consideration comments received: No.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units Nos. 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: December 13, 2001.
Brief description of amendments: The amendments revise Item d of
TS 5.5.11, ``Ventilation Filter Testing Program (VFTP),'' to lower
the maximum allowable differential pressure across the engineered
safety features ventilation systems units when tested at the
specified system flow rates.
Date of issuance: June 18, 2002.
Effective date: June 18, 2002, and shall be implemented within
60 days of the date of issuance.
Amendment Nos.: Unit 1-142, Unit 2-142, Unit 3-142.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 5, 2002 (67
FR 5325). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 18, 2002.
No significant hazards consideration comments received: No.
Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-318,
Calvert Cliffs Nuclear Power Plant, Unit No. 2, Calvert County,
Maryland
Date of application for amendment: November 19, 2001, as
supplemented March 27, 2002
Brief description of amendment: The amendment revises Technical
Specification 5.5.16 to eliminate the requirement to perform post-
modification containment integrated leakage rate testing following
replacement of the Unit 2 steam generators.
Date of issuance: June 27, 2002.
Effective date: As of the date of issuance to be implemented
following the Unit 2 refueling and steam generator replacement
outage in spring 2003.
Amendment No.: 230.
Renewed License No. DPR-69: Amendment revised the Technical
Specifications.
Date of initial notice in Federal Register: March 19, 2002 (67
FR 12599).
The March 27, 2002, supplemental letter provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation
dated June 27, 2002.
No significant hazards consideration comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: February 21, 2002.
Brief description of amendment: The amendment authorizes changes
to the Updated Final Safety Analysis Report (UFSAR) and the
Technical Requirements Manual to eliminate the chlorine detection
function from the control center heating, ventilation and air
conditioning system. Changes to the UFSAR are subject to the
requirements of 10 CFR 50.59; however, the changes were submitted to
the Nuclear Regulatory Commission for review and approval since they
involve the elimination of an automatic action.
Date of issuance: June 26, 2002.
Effective date: As of the date of issuance and shall be
implemented within 60 days.
Amendment No.: 147.
Facility Operating License No. NPF-43: Amendment revises the
UFSAR and TRM.
Date of initial notice in Federal Register: April 16, 2002 (67
FR 18643). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 26, 2002.
No significant hazards consideration comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: May 24, 2001.
Brief description of amendment: The amendment deletes License
Condition 2.C.(11), which required inspection of the low-pressure
turbine discs during the second refueling outage and specified that
the frequency of subsequent inspections should be in accordance with
the turbine manufacturer's recommendations. License Condition
2.C.(11) is no longer applicable to Fermi 2.
Date of issuance: June 26, 2002.
Effective date: As of the date of issuance and shall be
implemented within 30 days.
Amendment No.: 148.
Facility Operating License No. NPF-43: Amendment revises the
License.
Date of initial notice in Federal Register: December 12, 2001
(66 FR 64288). The Commission's related evaluation of the amendment
is contained in a Safety Evaluation dated June 26, 2002.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee County, South
Carolina
Date of application of amendments: June 21, 2000, as
supplemented by letters dated April 30 and May 20, 2002.
Brief description of amendments: The amendments authorize
changes to the Updated Final Safety Analysis Report Section 10.4.7,
``Emergency Feedwater System.''
Date of Issuance: June 11, 2002.
Effective date: As of the date of issuance and shall be
implemented within 30 days from the date of issuance.
Amendment Nos.: 325/325/326.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-
55: Amendments authorized changes to the UFSAR.
Date of initial notice in Federal Register: July 26, 2000 (65 FR
46008). The supplement dated April 30 and May 20, 2002, provided
clarifying information that did not change the scope of the June 21,
2000, application nor the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of
the amendments is contained in a Safety Evaluation dated June 11,
2002.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: April 16, 2001, as
supplemented by letters dated November 8, 2001, and February 11,
2002.
Brief description of amendment: The amendment authorizes the
licensee to modify the Final Safety Analysis Report (FSAR) to allow
an unisolable drain line between the reactor core isolation cooling
and the control rod drive/condensate pump rooms and identify the
pump room doors and penetration seals that are not watertight. In
addition, the change documents the minimum acceptable safe shutdown
equipment.
Date of issuance: June 19, 2002.
Effective date: June 19, 2002, and shall be implemented in the
next periodic update to the FSAR in accordance with 10 CFR 50.71(e).
Amendment No.: 176.
Facility Operating License No. NPF-21: The amendment revises the
FSAR.
Date of initial notice in Federal Register: May 16, 2001 (66 FR
27175). The November 8, 2001 and February 11, 2002, supplemental
letters provided additional information that clarified the
application, did not expand the scope of the application as
originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety
Evaluation dated June 19, 2002.
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 11, 2002.
Brief description of amendment: The amendment revises Technical
Specification Surveillance Requirement (SR) 3.0.3 to extend the
delay period, before entering a Limiting Condition for Operation,
following a missed Surveillance. The delay period is extended from
the current limit of ``* * * up to 24 hours or up to the limit of
the specified Frequency, whichever is less'' to ``* * * up to 24
hours or up to the limit of the specified
[[Page 45575]]
Frequency, whichever is greater.'' In addition, the following
requirement is added to SR 3.0.3: ``A risk evaluation shall be
performed for any Surveillance delayed greater than 24 hours and the
risk impact shall be managed.''
Date of issuance: June 27, 2002.
Effective date: June 27, 2002.
Amendment No.: 212.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 14, 2002 (67 FR
34485). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 27, 2002.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,
Unit No. 1, Pope County, Arkansas
Date of amendment request: March 13, 2002.
Brief description of amendment: The amendment revises
Surveillance Requirement (SR) 3.0.3 to extend the delay period
before entering a Limiting Condition for Operation, following a
missed surveillance. The delay period is extended from the current
limit of ``* * * up to 24 hours or up to the limit of the specified
Frequency, whichever is less'' to ``* * * up to 24 hours or up to
the limit of the specified Frequency, whichever is greater.'' In
addition, the following requirement is added to SR 3.0.3: ``A risk
evaluation shall be performed for any Surveillance delayed greater
than 24 hours and the risk impact shall be managed.''
Date of issuance: June 10, 2002.
Effective date: As of the date of issuance and shall be
implemented in conjunction with the implementation of Amendment No.
215.
Amendment No.: 217.
Renewed Facility Operating License No. DPR-51: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: April 30, 2002 (67
FR 21287). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 10, 2002.
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: August 1, 2001.
Brief description of amendments: These amendments revise
Limerick Generating Station's Units 1 and 2 Technical Specifications
by deleting Section 6.4, ``Training.''
Date of issuance: June 14, 2002.
Effective date: As of date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 160/122.
Facility Operating License Nos. NPF-39 and NPF-85: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 31, 2001 (66
FR 55018). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 14, 2002.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: April 18, 2002.
Brief description of amendment: The proposed amendment revises
Surveillance Requirement (SR) 3.0.3 to extend the delay period,
before entering a Limiting Condition for Operation, following a
missed surveillance. The delay period is extended from the current
limit of ``* * * up to 24 hours or up to the limit of the specified
Frequency, whichever is less'' to ``* * * up to 24 hours or up to
the limit of the specified Frequency, whichever is greater.'' In
addition, the following requirement is added to SR 3.0.3: ``A risk
evaluation shall be performed for any Surveillance delayed greater
than 24 hours and the risk impact shall be managed.''
Date of issuance: June 26, 2002.
Effective date: As of the date of issuance and shall be
implemented within 60 days of issuance.
Amendment No.: 203.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 14, 2002 (67 FR
34487). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 26, 2002.
No significant hazards consideration comments received: No.
GPU Nuclear Inc., Docket No. 50-320, Three Mile Island Nuclear
Station, Unit 2, Dauphin County, Pennsylvania
Date of amendment request: February 8, 2002.
Brief description of amendment request: The amendment would
replace referenced control requirements for access to high radiation
areas with the actual requirements of 10 CFR Part 20, and would
replace the existing Three Mile Island Nuclear Station, Unit 2,
Technical Specifications (TS) Section 6.11 with the wording
contained in Three Mile Island Nuclear Station, Unit 1, TS Section
6.12.
Date of issuance: June 27, 2002.
Effective date: As of the date of issuance and shall be
implemented within 30 days.
Amendment No.: 58.
Facility Operating License No. DPR-73: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 2, 2002 (67 FR
15623). The Commission's related evaluation of the amendment is
contained in a safety evaluation dated June 27, 2002.
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: June 18, 2001, as
supplemented by letters dated January 30, and March 1, 2002.
Brief description of amendment: The amendment revises (1) the
reference point for reactor vessel level instrumentation
specifications to use instrument ``zero'' instead of ``top of active
fuel;'' (2) simplifies the safety limits and limiting safety system
settings to eliminate specifications that are unnecessary, outdated,
or redundant to other Technical Specifications (TSs); (3) changes
the reactor coolant system pressure safety limit from 1335 psig to
1332 psig to correct a minor calculation error; and (4) makes
corresponding TS Bases changes.
Date of issuance: June 11, 2002.
Effective date: As of the date of issuance and shall be
implemented within 60 days.
Amendment No.: 128.
Facility Operating License No. DPR-22: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 25, 2001 (66 FR
38764). The supplements dated January 30 and March 1, 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 amendment is contained in a Safety Evaluation dated June 11,
2002.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of application for amendments: January 10, 2002.
Brief description of amendments: The amendments revise
Surveillance Requirement (SR) 3.0.3 to extend the delay period
before entering a Limiting Condition for Operation following a
missed surveillance. The delay period is extended from the current
limit of ``* * * up to 24 hours or up to the limit of the specified
Frequency, whichever is less'' to ``* * * up to 24 hours or up to
the limit of the specified Frequency, whichever is greater.'' In
addition, the following requirement is added to SR 3.0.3: ``A risk
evaluation shall be performed for any Surveillance delayed greater
than 24 hours and the risk impact shall be managed.''
Date of issuance: June 19, 2002.
Effective date: June 19, 2002, shall be implemented within 30
days from the date of issuance.
Amendment Nos.: Unit 1-153; Unit 2-153.
Facilit Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 5, 2002 (67 FR
10014). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 19, 2002.
[[Page 45576]]
No significant hazards consideration comments received: No.
Southern California Edison Company, Docket Nos. 50-361 and 50-362,
San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment request: May 22, 2002, as supplemented by
letters dated June 10, and June 14, 2002.
Brief description of amendment: This amendment revises Technical
Specification (TS) TS 5.5.2.11.f.1.h, ``Steam Generator (SG) Tube
Surveillance Program,'' to more clearly delineate the scope of the
SG tube inspection required in the tubesheet region. This TS change
will apply only to Cycle 12 (Unit 2) and Cycle 11 (Unit 3)
operations.
Date of issuance: June 17, 2002.
Effective date: June 17, 2002, to be implemented within 30 days
from the date of issuance.
Amendment Nos.: Unit 2--189 ; Unit 3--180.
Facility Operating License Nos. NPF-10 and NPF-15: The
amendments revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes (67 FR 38150 dated May 31, 2002). The notice
provided an opportunity to submit comments on the Commission's
proposed no significant hazards consideration determination. No
comments have been received. The notice also provided for an
opportunity to request a hearing by July 1, 2002, but indicated that
if the Commission makes a final no significant hazards consideration
determination any such hearing would take place after issuance of
the amendment. The Commission's related evaluation of the amendment,
finding of exigent circumstances, consultation with the State of
California and final determination of no significant hazards
consideration are contained in a Safety Evaluation dated June 17,
2002. The June 10, and June 14, 2002, supplemental letters provided
additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did
not change the staff's original proposed no significant hazards
consideration determination.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche
Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell County,
Texas
Date of amendment request: March 25, 2002, as supplemented by
the letter dated April 23, 2002.
Brief description of amendments: The proposed change revised the
Technical Specification (TS) 3.7.3, ``Feedwater Isolation Valves
(FIVs) and Associated Bypass Valves,'' to adopt the NUREG-1431,
``Standard Technical Specifications for Westinghouse Plants,''
Revision 2 version of the specification. The requirements of revised
TS 3.7.3 added, among other things, operability and suitable
surveillance requirements for Feedwater Control Valves and
Associated Bypass Valves and allowed for the extended out-of-service
time for one or more FIVs. In addition, a footnote which allowed a
one-time extension for Condition A Completion Time, has been deleted
because it is no longer applicable.
Date of issuance: June 20, 2002.
Effective date: As of the date of issuance and shall be
implemented within 60 days from the date of issuance.
Amendment Nos.: NPF-87, Amendment No. 97 and NPF-89, Amendment
No. 97.
Facility Operating License Nos. NPF-87 and NPF-89: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: May 14, 2002 (67 FR
34492). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 20, 2002.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 1st day of July 2002.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Acting Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 02-16956 Filed 7-8-02; 8:45 am]
BILLING CODE 7590-01-P