[Federal Register Volume 70, Number 20 (Tuesday, February 1, 2005)]
[Notices]
[Pages 5233-5254]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-1574]
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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 section 189a.(2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and 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 January 7, 2005, through January 19, 2005.
The last biweekly notice was published on January 18, 2005 (70 FR
2886).
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. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility
[[Page 5234]]
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.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, 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 general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include 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/requestor to relief. A petitioner/
requestor who fails to satisfy 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.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, 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 by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. 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 it is requested that copies be
[[Page 5235]]
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 requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the 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 PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: September 15, 2004.
Description of amendment request: The proposed amendment would
delete requirements from the Technical Specifications (TSs) to maintain
hydrogen recombiners and hydrogen and oxygen monitors. A notice of
availability for this TS improvement using the consolidated line item
improvement process was published in the Federal Register on September
25, 2003 (68 FR 55416). Licensees were generally required to implement
upgrades as described in NUREG-0737, ``Clarification of TMI [Three Mile
Island] Action Plan Requirements,'' and Regulatory Guide (RG) 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
combustible gas control were imposed by order for many facilities and
were added to, or included, in the TSs for nuclear power reactors
currently licensed to operate. The revised Title 10 of the Code of
Federal Regulations (10 CFR) Section 50.44, ``Standards for combustible
gas control system in light-water-cooled power reactors,'' eliminated
the requirements for hydrogen recombiners and relaxed safety
classifications and licensee commitments to certain design and
qualification criteria for hydrogen and oxygen monitors.
The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice
of availability of a model no significant hazards consideration (NSHC)
determination for referencing in license amendment applications in the
Federal Register on September 25, 2003 (68 FR 55416). The licensee
affirmed the applicability of the model NSHC determination in its
application dated September 15, 2004.
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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The NRC has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen and oxygen monitors are no longer required to mitigate
design-basis accidents and, therefore, the hydrogen monitors do not
meet the definition of a safety-related component as defined in 10
CFR 50.2. RG 1.97 Category 1, is intended for key variables that
most directly indicate the accomplishment of a safety function for
design-basis accident events. The hydrogen and oxygen monitors no
longer meet the definition of Category 1 in RG 1.97. As part of the
rulemaking to revise 10 CFR 50.44 the NRC found that Category 3, as
defined in RG 1.97, is an appropriate categorization for the
hydrogen monitors because the monitors are required to diagnose the
course of beyond design-basis accidents. Also, as part of the
rulemaking to revise 10 CFR 50.44, the NRC found that Category 2, as
defined in RG 1.97, is an appropriate categorization for the oxygen
monitors, because the monitors are required to verify the status of
the inert containment.
The regulatory requirements for the hydrogen and oxygen monitors
can be relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3,
[classification of the oxygen monitors as Category 2,] and removal
of the hydrogen and oxygen monitors from TS will not prevent an
accident management strategy through the use of the severe accident
management guidelines, the emergency plan, the emergency operating
procedures, and site survey monitoring that support modification of
emergency plan protective action recommendations.
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen and oxygen monitor
requirements, including removal of these requirements from TS, does
not involve a significant increase in the probability or 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 the hydrogen recombiner requirements and
relaxation of the hydrogen and oxygen monitor requirements,
including removal of these requirements from TS, will not result in
any failure mode not previously analyzed. The hydrogen recombiner
and hydrogen and oxygen monitor equipment was intended to mitigate a
design-basis hydrogen release. The hydrogen recombiner and hydrogen
and oxygen monitor equipment are not considered accident precursors,
nor does their existence or elimination have any adverse impact on
the pre-accident state of the reactor core or post accident
confinement of radionuclides within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen and oxygen monitor requirements,
including removal of these requirements from TS, 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.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The NRC has found that this hydrogen
release is not risk-significant
[[Page 5236]]
because the design-basis LOCA hydrogen release does not contribute
to the conditional probability of a large release up to
approximately 24 hours after the onset of core damage.
Category 3 hydrogen monitors are adequate 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 intent of the
requirements established as a result of the TMI, Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Category 2 oxygen monitors are adequate to verify the status of
an inerted containment.
Therefore, this change does not involve a significant reduction
in the margin of safety. The intent of the requirements established
as a result of the TMI, Unit 2 accident can be adequately met
without reliance on safety-related oxygen monitors. Removal of
hydrogen and oxygen monitoring from TS will not result in a
significant reduction in their functionality, reliability, and
availability.
The NRC staff proposes to determine that the amendment request
involves NSHC.
Attorney for licensee: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60666.
NRC Section Chief: Gene Y. Suh.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendments request: December 16, 2004.
Description of amendments request: The requested change will delete
Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure
Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated December 16, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating letter
report of shutdown experience and operating statistics if the
equivalent data is submitted using an industry electronic database.
It also eliminates the TS reporting requirement for an annual
occupational radiation exposure report, which provides information
beyond that specified in NRC regulations. The proposed change
involves no changes to plant systems or accident analyses. As such,
the change is administrative in nature and does not affect
initiators of analyzed events or assumed mitigation of accidents or
transients. 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 a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve significance hazards consideration.
Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix,
Arizona 85072-2034.
NRC Section Chief: Robert A. Gramm.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: December 1, 2004.
Description of amendments request: The requested change will delete
Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure
Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated December 1, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating report
of shutdown experience and operating statistics if the equivalent
data is submitted using an industry electronic database. It also
eliminates the TS reporting requirement for an annual occupational
radiation exposure report, which provides information beyond that
specified in NRC regulations. The proposed change involves no
changes to plant systems or accident analyses. As such, the change
is administrative in nature and does not affect initiators of
analyzed events or assumed mitigation of accidents or transients.
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 a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve significance hazards consideration.
Attorney for licensee: James M. Petro, Jr., Esquire, Counsel,
Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor,
Baltimore, MD 21202.
NRC Section Chief: Richard J. Laufer.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: December 6, 2004.
Description of amendment request: The requested change will delete
[[Page 5237]]
Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure
Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated December 6, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the TSs reporting requirements to
provide a monthly operating report of shutdown experience and
operating statistics if the equivalent data is submitted using an
industry electronic database. It also eliminates the TS reporting
requirement for an annual occupational radiation exposure report,
which provides information beyond that specified in NRC regulations.
The proposed change involves no changes to plant systems or accident
analyses. As such, the change is administrative in nature and does
not affect initiators of analyzed events or assumed mitigation of
accidents or transients. 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 a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve significant hazards consideration.
Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB,
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279 .
NRC Section Chief: M. Kotzalas (Acting).
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of amendment request: June 6, 2004.
Description of amendment request: The proposed change would modify
the Millstone Power Station, Unit No. 2 Technical Specifications (TSs)
to extend the 10-year test interval for the Integrated Leakage Rate
Test program to 15 years from the last Type A test. Specifically, the
proposed change would revise TS 6.19, ``Containment Leakage Rate
Testing [CLRT] Program,'' and permit a one-time, 5-year extension of
the 10-year performance-based Type A test interval. In addition, the
testing would be in accordance with the CLRT Program, Regulatory Guide
(RG) 1.163, ``Performance-Based Containment Leak-Test Program'' and
surveillance testing requirements as proposed in Nuclear Energy
Institute 94-01 for Type A testing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The proposed extension to Type A testing cannot increase the
probability of an accident previously evaluated since extension of
the containment Type A testing is not a physical plant modification
that could alter the probability of accident occurrence, nor is it
an activity or modification that by itself could lead to equipment
failure or accident initiation.
The proposed one-time, five-year extension to Type A testing
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
C tests. It concludes that even reducing the Type A (ILRT
[integrated leak rate test]) testing frequency to once per twenty
years leads to an imperceptible increase in risk.
DNC (the licensee) provides a high degree of assurance through
indirect testing and inspection that the containment will not
degrade in a manner detectable only by Type A testing. The last two
Type A tests identified containment leakage within acceptance
criteria, indicating a very leak-tight containment. Inspections
required by the ASME Code [American Society of Mechanical Engineers
Boiler and Pressure Vessel Code] are also performed in order to
identify indications of containment degradation that could affect
leak-tightness. Separately, Type B and C testing required by
Technical Specifications, identifies any containment opening from
design penetrations, such as valves, that would otherwise be
detected by a Type A test. These factors establish that a one-time,
five-year extension to the Millstone Unit 2 Type A test interval
will not represent a significant increase in the consequences of an
accident.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
The proposed revision to the Technical Specifications adds a
one-time extension to the current interval for Type A testing for
Millstone Unit 2. 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
testing does not create the possibility of a new or different type
of accident since there are no physical changes being made to the
plant and there are no changes to the operation of the plant that
could introduce a new failure.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
The proposed revision to Millstone Unit 2 Technical
Specifications adds 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 for Millstone Unit 2. RG 1.174
provides guidance for determining the risk impact of plant-specific
changes to the licensing basis. RG 1.174 defines very small changes
in risk as resulting in increases of CDF [core damage frequency]
below 10-\6\/yr and increases in LERF [large early
release frequency] below 10-\7\/yr. Since the ILRT does
not impact CDF, the relevant criterion is LERF. The increase in
LERF, resulting from a change in the Type A ILRT test interval from
a once-per-ten-years to a once-per-fifteen-years is 0.83 x
10-\8\/yr, based on internal events. Since guidance in
Reg. Guide 1.174 defines very small changes in LERF as below
10-\7\/yr, increasing the ILRT interval from ten to
fifteen years is, therefore, considered non-risk significant and
will not significantly reduce the margin of safety. The NUREG-1493
generic study of the effects of extending containment leakage
testing found that a 20-year interval in Type A leakage testing
resulted in an imperceptible increase in risk to the public. NUREG-
1493 generically concludes that the design containment leakage rate
contributes about 0.1 percent of the overall risk. Decreasing the
Type A testing frequency would have a minimal effect on this risk
since 95% of the Type A detectable leakage paths would already be
detected by Type B and C testing. Given that the proposed change
will continue to meet the current design basis, any reduction in a
margin of safety would not be significant.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three
[[Page 5238]]
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, Waterford, CT
06385.
NRC Section Chief: Darrell J. Roberts.
Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone
Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: December 16, 2004.
Description of amendment request: The proposed amendment would
revise the current fuel rod average licensing basis burnup limit for
one lead test assembly (LTA) containing advanced zirconium based alloys
to a limit not exceeding 71,000 MWD/MTU.
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. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The Westinghouse LTA is very similar in design to the
Westinghouse fuel that comprises the remainder of the core. The
reload core design for Millstone Unit 3 Cycle 12, where one LTA will
operate to high burnup, will meet all applicable design criteria.
The performance of the Emergency Core Cooling System will not be
affected by the operation of the LTA and operation of the LTA to
high burnup will not result in a change to the Millstone Unit 3
reload design and safety analysis limits. Operation of one
Westinghouse LTA to high burnup will not result in a measurable
impact on normal operating releases, and will not increase the
predicted radiological consequences of accidents postulated in
Chapter 15 of the Millstone FSAR [final safety analysis report].
Therefore, neither the probability of occurrence nor the
consequences of any accident previously evaluated is significantly
increased.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The Westinghouse LTA is very similar in design (both mechanical
and composition of materials) to the resident Westinghouse fuel. All
design and performance criteria will continue to be met and no new
single failure mechanisms will be created. The irradiation of one
LTA to high burnup does not involve any alteration to plant
equipment or procedures, which would introduce any new or unique
operational modes or accident precursors. Therefore, the possibility
for a new or different kind of accident from any accident previously
evaluated is not created.
3. Involve a significant reduction in a margin of safety.
The operation of one Westinghouse LTA to high burnup does not
change the performance requirements of any system or component such
that any design criteria will be exceeded. The normal limits on core
operation defined in the Millstone Unit 3 Technical Specifications
will remain applicable for the core in which the high burnup
assembly is irradiated. Therefore, the margin of safety as defined
in the Bases to the Millstone Unit 3 Technical Specifications is not
significantly reduced.
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: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
NRC Section Chief: Darrell Roberts.
Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423,
Millstone Power Station, Unit Nos. 2 and 3, New London County,
Connecticut
Date of amendment request: September 8, 2004.
Description of amendment request: The proposed amendment deletes
the requirements from the technical specifications (TSs) to maintain
hydrogen recombiners and hydrogen monitors. Licensees were generally
required to implement upgrades as described in NUREG-0737,
``Clarification of TMI [Three Mile Island] Action Plan Requirements,''
and Regulatory Guide (RG) 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 combustible gas control were
imposed by Order for many facilities and were added to or included in
the TSs for nuclear power reactors currently licensed to operate. The
revised 10 CFR 50.44, ``Standards for Combustible Gas Control System in
Light-Water-Cooled Power Reactors,'' eliminated the requirements for
hydrogen recombiners and relaxed safety classifications and licensee
commitments to certain design and qualification criteria for hydrogen
and oxygen monitors.
The NRC staff issued a notice of availability of a model no
significant hazards consideration determination for referencing in
license amendment applications in the Federal Register on September 25,
2003 (68 FR 55416). The licensee affirmed the applicability of the
model NSHC determination in its application dated September 8, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen monitors are no longer required to mitigate design-basis
accidents and, therefore, the hydrogen monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
Category 1 in RG 1.97 is intended for key variables that most
directly indicate the accomplishment of a safety function for
design-basis accident events. The hydrogen monitors no longer meet
the definition of Category 1 in RG 1.97. As part of the rulemaking
to revise 10 CFR 50.44 the Commission found that Category 3, as
defined in RG 1.97, is an appropriate categorization for the
hydrogen monitors because the monitors are required to diagnose the
course of beyond design-basis accidents.
The regulatory requirements for the hydrogen monitors can be
relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3, and removal
of the hydrogen monitors from TS will not prevent an accident
management strategy through the use of the severe accident
management guidelines (SAMGs), the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
[[Page 5239]]
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen monitor requirements,
including removal of these requirements from TSs, does not involve a
significant increase in the probability or 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 the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TSs, will not result in any failure mode
not previously analyzed. The hydrogen recombiner and hydrogen
monitor equipment was intended to mitigate a design-basis hydrogen
release. The hydrogen recombiner and hydrogen monitor equipment are
not considered accident precursors, nor does their existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post-accident confinement of radionuclides within
the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TSs, 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.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate 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 intent of the
requirements established as a result of the TMI Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Therefore, this change does not involve a significant reduction
in the margin of safety. Removal of hydrogen monitoring from TSs
will not result in a significant reduction in their functionality,
reliability, and availability.
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, Waterford, CT
06385.
NRC Section Chief: Darrell J. Roberts.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina; Docket Nos.
50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3,
Oconee County, South Carolina
Date of amendment request: September 20, 2004.
Description of amendment request: The proposed amendment deletes
the requirements from the technical specifications (TS) to maintain
hydrogen recombiners (McGuire only) and hydrogen monitors (McGuire and
Oconee). Licensees were generally required to implement 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
combustible gas control were imposed by Order for many facilities and
were added to or included in the TS for nuclear power reactors
currently licensed to operate. The revised 10 CFR 50.44, ``Standards
for Combustible Gas Control System in Light-Water-Cooled Power
Reactors,'' eliminated the requirements for hydrogen recombiners and
relaxed safety classifications and licensee commitments to certain
design and qualification criteria for hydrogen and oxygen monitors.
The NRC staff issued a notice of availability of a model no
significant hazards consideration determination for referencing in
license amendment applications in the Federal Register on September 25,
2003 (68 FR 55416). The licensee affirmed the applicability of the
model NSHC determination in its application dated September 20, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen monitors are no longer required to mitigate design-basis
accidents and, therefore, the hydrogen monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
Category 1 in [Regulatory Guide] RG 1.97 is intended for key
variables that most directly indicate the accomplishment of a safety
function for design-basis accident events. The hydrogen monitors no
longer meet the definition of Category 1 in RG 1.97. As part of the
rulemaking to revise 10 CFR 50.44 the Commission found that Category
3, as defined in RG 1.97, is an appropriate categorization for the
hydrogen monitors because the monitors are required to diagnose the
course of beyond design-basis accidents.
The regulatory requirements for the hydrogen monitors can be
relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3, and removal
of the hydrogen monitors from [Technical Specification] TS will not
prevent an accident management strategy through the use of the
severe accident management guidelines (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 the hydrogen recombiner
requirements and relaxation of the hydrogen monitor requirements,
including removal of these requirements from TS, does not involve a
significant increase in the probability or the consequences of any
accident previously evaluated.
[[Page 5240]]
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 the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, will not result in any failure mode
not previously analyzed. The hydrogen recombiner and hydrogen
monitor equipment was intended to mitigate a design-basis hydrogen
release. The hydrogen recombiner and hydrogen monitor equipment are
not considered accident precursors, nor does their existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within
the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, 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.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate 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 intent of the
requirements established as a result of the TMI Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Therefore, this change does not involve a significant reduction
in the margin of safety. Removal of hydrogen monitoring from TS will
not result in a significant reduction in their functionality,
reliability, and availability.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
Entergy Nuclear Operations, Docket Nos. 50-247 and 50-286, Indian Point
Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York
Date of amendment request: October 22, 2004.
Description of amendment request: The proposed amendments would
delete the requirements from the Technical Specifications (TSs) to
maintain hydrogen recombiners and hydrogen monitors. Licensees were
generally required to implement upgrades as described in NUREG-0737,
``Clarification of TMI [Three Mile Island] Action Plan Requirements,''
and Regulatory Guide (RG) 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 combustible gas control were
imposed by Order for many facilities and were added to or included in
the TSs for nuclear power reactors currently licensed to operate. The
revised 10 CFR 50.44, ``Standards for Combustible Gas Control System in
Light-Water-Cooled Power Reactors,'' eliminated the requirements for
hydrogen recombiners and relaxed safety classifications and licensee
commitments to certain design and qualification criteria for hydrogen
and oxygen monitors.
The Nuclear Regulatory Commission (NRC) staff issued a notice of
availability of a model no significant hazards consideration (NSHC)
determination for referencing in license amendment applications in the
Federal Register on September 25, 2003 (68 FR 55416). The licensee
affirmed the applicability of the model NSHC determination in its
application dated October 22, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen monitors are no longer required to mitigate design-basis
accidents and, therefore, the hydrogen monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
Category 1 in RG 1.97 is intended for key variables that most
directly indicate the accomplishment of a safety function for
design-basis accident events. The hydrogen monitors no longer meet
the definition of Category 1 in RG 1.97. As part of the rulemaking
to revise 10 CFR 50.44 the Commission found that Category 3, as
defined in RG 1.97, is an appropriate categorization for the
hydrogen monitors because the monitors are required to diagnose the
course of beyond design-basis accidents.
The regulatory requirements for the hydrogen monitors can be
relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3, and removal
of the hydrogen monitors from TS will not prevent an accident
management strategy through the use of the severe accident
management guidelines (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 the hydrogen recombiner
requirements and relaxation of the hydrogen monitor requirements,
including removal of these requirements from TS, does not involve a
significant increase in the probability or 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 the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, will not result in any failure mode
not previously analyzed. The hydrogen recombiner and hydrogen
monitor equipment was intended to mitigate a design-basis hydrogen
release. The hydrogen recombiner and hydrogen monitor equipment are
not considered accident precursors, nor does their existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement
[[Page 5241]]
of radionuclides within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, 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.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate 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 intent of the
requirements established as a result of the TMI Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Therefore, this change does not involve a significant reduction
in the margin of safety. Removal of hydrogen monitoring from TS will
not result in a significant reduction in their functionality,
reliability, and availability.
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, Docket Nos. 50-247 and 50-286, Indian Point
Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York
Date of amendment request: October 25, 2004.
Description of amendment request: The requested change will delete
Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure
Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated October 25, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating report
of shutdown experience and operating statistics if the equivalent
data is submitted using an industry electronic database. It also
eliminates the TS reporting requirement for an annual occupational
radiation exposure report, which provides information beyond that
specified in NRC regulations. The proposed change involves no
changes to plant systems or accident analyses. As such, the change
is administrative in nature and does not affect initiators of
analyzed events or assumed mitigation of accidents or transients.
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 a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in [a] margin of safety.
Based upon the reasoning presented above, the requested change does
not involve significance 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: December 30, 2004.
Description of amendment request: The proposed amendment would
revise a Technical Specification (TS) surveillance requirement (SR) in
TS 3.1.4, ``Control Rod Scram Times.'' Specifically, the proposed
change would revise the frequency for SR 3.1.4.2, ``Control Rod Scram
Time Testing,'' from ``120 days cumulative operation in MODE 1'' to
``200 days cumulative operation in MODE 1.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in licensing amendment applications in the Federal Register on August
23, 2004 (69 FR 51864). The licensee affirmed the applicability of the
model NSHC determination in its application dated December 30, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change extends the frequency for testing control
rod scram time testing from every 120 days of cumulative Mode 1
operation to 200 days of cumulative Mode 1 operation. The frequency
of surveillance testing is not an initiator of any accident
previously evaluated. The frequency of surveillance testing does not
affect the ability to mitigate any accident previously evaluated, as
the tested component is still required to be operable. Therefore,
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The proposed change extends the frequency for testing control
rod scram time testing from every 120 days of cumulative Mode 1
operation to 200 days of cumulative Mode 1 operation. The proposed
change does not result in any new or different modes of plant
operation. 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 extends the frequency for testing control
rod scram time
[[Page 5242]]
testing from every 120 days of cumulative Mode 1 operation to 200
days of cumulative Mode 1 operation. The proposed change continues
to test the control rod scram time to ensure the assumptions in the
safety analysis are protected. Therefore, the proposed change does
not involve a significant reduction in a margin of safety.
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 Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: December 20, 2004.
Description of amendment request: The proposed amendment would
increase the lifting tripod's rating from 150 tons to 190 tons. This
would allow for additional flexibility when lifting the new reactor
vessel head during refueling outages.
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 ANO-1 [Arkansas Nuclear One, Unit 1] Tripod does not perform
a safety function required by 10 CFR [Part] 50. The Tripod serves to
perform heavy load movements during refueling outages[,] including
[movement of] the reactor vessel head. Safe load paths have been
established in accordance with NUREG-0612[, ``Control of Heavy Loads
at Nuclear Power Plants,''] to ensure that the fuel and safety[-
]related equipment required to be inservice are protected. Use of
actual Tripod eyelet Certified Material Test Reports (CMTRs)
demonstrates that a safety factor of 3 to yield is maintained and
that the lifting devices will perform their design function under
maximum lifted loads. The Tripod does not serve any mitigative
functions to lessen accidents.
Therefore, the proposed change does not affect the probability
or consequences of any ANO-1 analyzed accidents.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The only time that the Tripod is performing heavy loads
movements is during Refueling operations. Safe load paths and load
drop analyses have been performed to assure that heavy loads
movements will not cause fuel damage or cause safety[-]related
equipment to become inoperable. The proposed use of CMTRs instead of
minimum yield strength of the material still assures that the Tripod
will perform its required function to not create an accident. In
addition, there is no change to the operation of the Tripod that
would create a new failure mode or possible accident.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The design margin for the Tripod is established by NUREG-0612
and ANSI [American National Standards Institute] N14.6-1978[,
``Special Lifting Devices for Shipping Containers Weighing 10,000
Pounds or More for Nuclear Materials'']. A factor of safety of 3 for
yield strength and 5 for ultimate strength for both the static and
dynamic load factors is required to be met. These factors of safety
provide sufficient margin to assure that the Tripod will perform its
design function of maximum lifted loads. In addition, the use [of] a
dynamic load factor of 1.15 above the static load is well above the
actual dynamic factor to be experienced from the design lift speed
of the polar crane. The use of CMTRs does not result in a
significant reduction in the margin of safety of the Tripod. In
addition, the Tripod will be load tested to 150% [percent] of its
design static and dynamic loading which will further assure adequate
safety margin.
Therefore, the margin of safety is not changed by the proposed
change to the ANO-1 SAR [Safety Analysis Report].
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: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: December 17, 2004.
Description of amendment request: The proposed change will revise
the air lock surveillance test acceptance criteria to be consistent
with the NRC approved Industry Technical Specification Task Force
(TSTF) change to the Standard Technical Specifications (STS), TSTF-52,
entitled ``Implement 10 CFR [Part] 50, Appendix J, Option B.'' By
letter dated April 6, 1998, the NRC Staff issued amendment number 135
to the GGNS license permitting the implementation of the containment
leak rate testing provisions of 10 CFR Part 50, Appendix J, Option B.
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.
Primary containment air lock leak rate testing can have no
effect on the probability of any postulated accident. The proposed
change will increase the allowed containment air lock leakage rate
and convert it from an absolute leakage rate to a percentage of the
overall primary containment leakage rate. No change to the overall
leakage rate of the containment is being proposed, therefore there
is no change to the consequences of any postulated accident. The
change in air lock leakage rate will not impact the design or
operation of any plant system or component nor will they affect
initiation or mitigation of any accidents previously analyzed.
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 primary containment air locks form part of the primary
containment pressure boundary. The periodic containment air lock
leakage rate tests specified in SR 3.6.1.2.1 verifies that the air
lock leakage does not exceed the allowed fraction of the overall
primary containment leakage rate. This request involves a change in
the allowable leakage rate of the primary containment air locks
without increasing the overall allowed leakage rate of the
containment. Changing the allowable leakage rate has no influence
on, nor does it contribute in any way to, the possibility of a new
or different kind of accident or malfunction from those previously
analyzed. There will be no effect on the types and amounts of
overall leakage from the primary containment boundary. The proposed
amendment will not produce any changes to the design or operation of
the plant. The method of performing the test is not changed. No new
accident modes are created by changing the allowable leakage in this
manner. No safety-related equipment or safety functions are altered
as a result of this 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?
[[Page 5243]]
Response: No.
Air lock integrity and leak tightness are essential for
maintaining primary containment leakage rate to within limits in the
event of a design basis accident. The periodic containment air lock
leakage rate tests verify that the air lock leakage does not exceed
the allowed fraction of the overall primary containment leakage
rate. Since no changes are proposed to the maximum allowable primary
containment leakage rate, the design basis radiological analysis is
not impacted by this change. The license amendment request removes
unnecessary conservatism from the testing program and allows
consistency with current industry practice. Since no changes are
proposed to the maximum allowable primary containment leakage rate,
the design basis radiological analysis is not impacted by this
change.
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: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
NRC Section Chief: Michael K. Webb.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois; Docket Nos. 50-237 and 50-249, Dresden Nuclear Power
Station, Units 2 and 3, Grundy County, Illinois; Docket Nos. 50-373 and
50-374, LaSalle County Station, Units 1 and 2, LaSalle County,
Illinois; Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power
Station, Units 1 and 2, Rock Island County, Illinois
Date of amendment request: September 15, 2004.
Description of amendment request: The proposed amendment would
delete requirements from the Technical Specifications (TSs) to maintain
hydrogen recombiners and hydrogen and oxygen monitors. Licensees were
generally required to implement upgrades as described in NUREG-0737,
``Clarification of TMI [Three Mile Island] Action Plan Requirements,''
and Regulatory Guide (RG) 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 combustible gas control were
imposed by order for many facilities and were added to, or included, in
the TSs for nuclear power reactors currently licensed to operate. The
revised Title 10 of the Code of Federal Regulations (10 CFR) Section
50.44, `` Combustible gas control for nuclear power reactors,''
eliminated the requirements for hydrogen recombiners and relaxed safety
classifications and licensee commitments to certain design and
qualification criteria for hydrogen and oxygen monitors.
The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice
of availability of a model no significant hazards consideration (NSHC)
determination for referencing in license amendment applications in the
Federal Register on September 25, 2003 (68 FR 55416). The licensee
affirmed the applicability of the model NSHC determination in its
application dated September 15, 2004.
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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen and oxygen monitors are no longer required to mitigate
design-basis accidents and, therefore, the hydrogen monitors do not
meet the definition of a safety-related component as defined in 10
CFR 50.2. RG 1.97 Category 1, is intended for key variables that
most directly indicate the accomplishment of a safety function for
design-basis accident events. The hydrogen and oxygen monitors no
longer meet the definition of Category 1 in RG 1.97. As part of the
rulemaking to revise 10 CFR 50.44 the Commission found that Category
3, as defined in RG 1.97, is an appropriate categorization for the
hydrogen monitors because the monitors are required to diagnose the
course of beyond design-basis accidents. Also, as part of the
rulemaking to revise 10 CFR 50.44, the Commission found that
Category 2, as defined in RG 1.97, is an appropriate categorization
for the oxygen monitors, because the monitors are required to verify
the status of the inert containment.
The regulatory requirements for the hydrogen and oxygen monitors
can be relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3,
classification of the oxygen monitors as Category 2, and removal of
the hydrogen and oxygen monitors from TS will not prevent an
accident management strategy through the use of the SAMGs [severe
accident management guidelines], the emergency plan (EP), the
emergency operating procedures (EOPs), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen and oxygen monitor
requirements, including removal of these requirements from TS, does
not involve a significant increase in the probability or 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 the hydrogen recombiner requirements and
relaxation of the hydrogen and oxygen monitor requirements,
including removal of these requirements from TS, will not result in
any failure mode not previously analyzed. The hydrogen recombiner
and hydrogen and oxygen monitor equipment was intended to mitigate a
design-basis hydrogen release. The hydrogen recombiner and hydrogen
and oxygen monitor equipment are not considered accident precursors,
nor does their existence or elimination have any adverse impact on
the pre-accident state of the reactor core or post accident
confinement of radionuclides within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen and oxygen monitor requirements,
including removal of these requirements from TS, 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.
[[Page 5244]]
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate 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 intent of the
requirements established as a result of the TMI, Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Category 2 oxygen monitors are adequate to verify the status of
an inerted containment.
Therefore, this change does not involve a significant reduction
in the margin of safety. The intent of the requirements established
as a result of the TMI, Unit 2 accident can be adequately met
without reliance on safety-related oxygen monitors. Removal of
hydrogen and oxygen monitoring from TS will not result in a
significant reduction in their functionality, reliability, and
availability.
Based on 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: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Gene Y. Suh.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station (DNPS), Units 2 and 3, Grundy County, Illinois;
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station (QCNPS), Units 1 and 2, Rock Island
County, Illinois
Date of amendment request: November 4, 2004.
Description of amendment request: The proposed amendments would
revise the plant technical specification (TS) pressure and temperature
(P/T) limit curves for 54 effective full power years (EFPY) to support
a 20-year license extension for both DNPS and QCNPS to 60 years (i.e.,
54 EFPY), and resolves a non-conservative condition for TS Section
3.4.9, Figure 3.4.9-2, ``Non-Nuclear Heatup/Cooldown Curve,'' for
QCNPS.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR) section 50.91(a), Exelon Generation Company (EGC)
has provided its analysis of the issue of no significant hazards
consideration (NSHC), which is presented below:
According to 10 CFR 50.92, ``Issuance of amendment,'' paragraph
(c), a proposed amendment to an operating license involves no
significant hazards consideration if operation of the facility in
accordance with the proposed amendment would not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated; or
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated; or
(3) Involve a significant reduction in a margin of safety.
In support of this determination, an evaluation of each of the
three criteria set forth in 10 CFR 50.92 is provided below regarding
the proposed license amendment.
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes request that, for DNPS, Units 2 and 3 and
QCNPS, Units 1 and 2, P/T limit curves in TS 3.4.9, ``RCS Pressure
and Temperature (P/T) Limits,'' be revised.
The P/T limits are prescribed during all operational conditions
to avoid encountering pressure, temperature, and temperature rate-
of-change conditions that might cause undetected flaws to propagate,
resulting in non-ductile failure of the reactor coolant pressure
boundary, which is an unanalyzed condition. The methodology used to
determine the P/T limits has been approved by the NRC [Nuclear
Regulatory Commission] and thus is an acceptable method for
determining these limits. Therefore, the proposed changes do not
affect the probability of an accident previously evaluated.
There is no specific accident that postulates a non-ductile
failure of the reactor coolant pressure (RCP) boundary. The loss of
coolant accident analyzed for the plant assumes a 4.281 square feet
complete break of the recirculation pump suction line. The revision
to the P/T limits does not change this assumption. Thus, the
radiological consequences of any accident previously evaluated are
not increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not change the response of plant
equipment to transient conditions. The proposed changes do not
introduce any new equipment, modes of system operation, or failure
mechanisms.
Non-ductile failure of the RCP boundary is not an analyzed
accident. The proposed changes to the P/T limits were developed
using an NRC-approved methodology, and thus the revised limits will
continue to provide protection against non-ductile failure of the
RCP boundary.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
The margin of safety related to the proposed changes is the
margin between the proposed P/T limits and the pressures and
temperatures that would produce nonductile failure of the RCP
boundary. NRC requirements to protect the integrity of the reactor
coolant pressure boundary in nuclear power plants is established in
10 CFR 50, Appendix G, ``Fracture Toughness Requirements,'' which
requires that the P/T limits for an operating plant be at least as
conservative as those that would be generated if the methods of
American Society of Mechanical Engineers, Section XI, Appendix G,
were applied. The use of an NRC-approved methodology, together with
conservatively chosen plant-specific input parameters, provides an
acceptable margin of safety. Therefore, the proposed changes do not
involve a significant reduction in a margin of safety.
Based upon the above responses, EGC concluded that the proposed
amendment presents no significant hazards consideration under the
standards set forth in 10 CFR 50.92 and, accordingly, a finding of
no significant hazards consideration is justified.
The NRC staff has reviewed EGC'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 NSHC.
Attorney for licensee: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Gene Y. Suh.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of application for amendments: September 15, 2004.
Description of amendment request: The proposed amendment would
delete requirements from the Technical Specifications (TSs) to maintain
containment hydrogen and oxygen monitors. A notice of availability for
this technical specification improvement using the consolidated line
item improvement process (CLIIP) was published in the Federal Register
on September 25, 2003 (68 FR 55416). Licensees were generally required
to implement 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.''
[[Page 5245]]
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI, Unit 2. Requirements related to
combustible gas control were imposed by Order for many facilities and
were added to or included in the TSs for nuclear power reactors
currently licensed to operate. The revised 10 CFR 50.44, ``Standards
for combustible gas control system in light-water-cooled power
reactors,'' eliminated the requirements for hydrogen recombiners and
relaxed safety classifications and licensee commitments to certain
design and qualification criteria for hydrogen and oxygen monitors.
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on September
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the
relevant portions of the model NSHC determination (hydrogen and oxygen
monitors only) in its application dated September 15, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen and oxygen monitors are no longer required to mitigate
design-basis accidents and, therefore, the hydrogen monitors do not
meet the definition of a safety-related component as defined in 10
CFR 50.2. RG [Regulatory Guide] 1.97 Category 1, is intended for key
variables that most directly indicate the accomplishment of a safety
function for design-basis accident events. The hydrogen and oxygen
monitors no longer meet the definition of Category 1 in RG 1.97. As
part of the rulemaking to revise 10 CFR 50.44 the Commission found
that Category 3, as defined in RG 1.97, is an appropriate
categorization for the hydrogen monitors because the monitors are
required to diagnose the course of beyond design-basis accidents.
Also, as part of the rulemaking to revise 10 CFR 50.44, the
Commission found that Category 2, as defined in RG 1.97, is an
appropriate categorization for the oxygen monitors, because the
monitors are required to verify the status of the inert containment.
The regulatory requirements for the hydrogen and oxygen monitors
can be relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3,
[classification of the oxygen monitors as Category 2,] and removal
of the hydrogen and oxygen monitors from TS will not prevent an
accident management strategy through the use of the severe accident
management guidelines (SAMGs), the emergency plan (EP), the
emergency operating procedures (EOPs), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen and oxygen monitor
requirements, including removal of these requirements from TS, does
not involve a significant increase in the probability or 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 the hydrogen recombiner requirements and
relaxation of the hydrogen and oxygen monitor requirements,
including removal of these requirements from TS, will not result in
any failure mode not previously analyzed. The hydrogen recombiner
and hydrogen and oxygen monitor equipment was intended to mitigate a
design-basis hydrogen release. The hydrogen recombiner and hydrogen
and oxygen monitor equipment are not considered accident precursors,
nor does their existence or elimination have any adverse impact on
the pre-accident state of the reactor core or post accident
confinement of radionuclides within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen and oxygen monitor requirements,
including removal of these requirements from TS, 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.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate 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 intent of the
requirements established as a result of the TMI, Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Category 2 oxygen monitors are adequate to verify the status of
an inerted containment.
Therefore, this change does not involve a significant reduction
in the margin of safety. The intent of the requirements established
as a result of the TMI, Unit 2 accident can be adequately met
without reliance on safety-related oxygen monitors. Removal of
hydrogen and oxygen monitoring from TS will not result in a
significant reduction in their functionality, reliability, and
availability.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for Licensee: Thomas S. O'Neill, Associate and General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Darrell Roberts.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: September 21, 2004.
Description of amendment request: The proposed amendment deletes
the requirements from the technical specifications (TS) to maintain
containment hydrogen monitors. Licensees were generally required to
implement upgrades as described in NUREG-0737, ``Clarification of TMI
[Three Mile Island] Action Plan Requirements,'' and Regulatory Guide
(RG) 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 combustible
[[Page 5246]]
gas control were imposed by Order for many facilities and were added to
or included in the TS for nuclear power reactors currently licensed to
operate. The revised 10 CFR 50.44, ``Standards for Combustible Gas
Control System in Light-Water-Cooled Power Reactors,'' eliminated the
requirements for hydrogen recombiners and relaxed safety
classifications and licensee commitments to certain design and
qualification criteria for hydrogen and oxygen monitors.
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on September
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the
relevant portions of the model NSHC determination (hydrogen monitors
only) in its application dated September 21, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen monitors are no longer required to mitigate design-basis
accidents and, therefore, the hydrogen monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
Category 1 in RG 1.97 is intended for key variables that most
directly indicate the accomplishment of a safety function for
design-basis accident events. The hydrogen monitors no longer meet
the definition of Category 1 in RG 1.97. As part of the rulemaking
to revise 10 CFR 50.44 the Commission found that Category 3, as
defined in RG 1.97, is an appropriate categorization for the
hydrogen monitors because the monitors are required to diagnose the
course of beyond design-basis accidents.
The regulatory requirements for the hydrogen monitors can be
relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3, and removal
of the hydrogen monitors from TS will not prevent an accident
management strategy through the use of the severe accident
management guidelines (SAMGs), the emergency plan (EP), the
emergency operating procedures (EOPs), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen monitor requirements,
including removal of these requirements from TS, does not involve a
significant increase in the probability or 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 the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, will not result in any failure mode
not previously analyzed. The hydrogen recombiner and hydrogen
monitor equipment was intended to mitigate a design-basis hydrogen
release. The hydrogen recombiner and hydrogen monitor equipment are
not considered accident precursors, nor does their existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within
the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, 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.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate 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 intent of the
requirements established as a result of the TMI Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Therefore, this change does not involve a significant reduction
in the margin of safety. Removal of hydrogen monitoring from TS will
not result in a significant reduction in their functionality,
reliability, and availability.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Michael L. Marshall.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: January 13, 2005.
Description of amendment request: The proposed change would allow a
one-time extended allowed outage time (AOT) change to Improved
Technical Specifications (ITS) 3.5.2, Emergency Core Cooling Systems
(ECCS)--Operating; 3.6.6, Reactor Building Spray and Containment
Cooling Systems; 3.7.8, Decay Heat Closed Cycle Cooling Water System
(DC); and 3.7.10, Decay Heat Seawater System to allow the refurbishment
of Decay Heat Seawater System Pump RWP-3B online.
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:
This request has been evaluated against the standards in 10 CFR
50.92, and has been determined to not involve a significant hazards
consideration. In support of this conclusion, the following analysis
is provided:
1. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed license amendment extends, on a one-time basis, the
Completion Time for the systems described above from 72 hours to 10
days. These Systems are designed to provide cooling for components
essential to the mitigation of plant transients and
[[Page 5247]]
accidents. The systems are not initiators of design basis accidents.
The proposed ITS changes have been evaluated to assess their impact
on normal operation of the systems affected and to ensure that their
design basis safety functions are preserved.
A Probabilistic Safety Assessment (PSA) has been performed to
assess the risk impact of an increase in Completion Time from 72
hours to 10 days. Although the proposed one-time change results in
an increase in Core Damage Frequency (CDF) and Large Early Release
Frequency (LERF), the value of these increases are considered as
small (CDF) and very small (LERF) in the current regulatory
guidance.
Therefore, granting this LAR [License Amendment Request] does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does not create the possibility of a new or different type of
accident from any accident previously evaluated.
The proposed license amendment extends, on a one-time basis, the
Completion Time for the systems described above from 72 hours to 10
days.
The proposed LAR will not result in changes to the design,
physical configuration of the plant or the assumptions made in the
safety analysis. Therefore, the proposed change will not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does not involve a significant reduction in the margin of
safety.
The proposed license amendment extends, on a one-time basis, the
Completion Time for the systems described above from 72 hours to 10
days. The proposed change will allow online repair of Decay Heat
Seawater pump RWP-3B to restore the pump to full qualification which
will improve its reliability and useful lifetime, thus increasing
the long term margin of safety of the system.
The proposed LAR will reduce the probability (and associated
risk) of a plant shutdown to repair a Decay Heat Services Seawater
pump. To ensure defense-in-depth capabilities and the assumptions in
the risk assessment are maintained during the proposed one-time
extended Completion Time, CR-3 will continue the performance of 10
CFR 50.65(a)(4) assessments before performing maintenance or
surveillance activities and no maintenance activities of other risk
sensitive equipment beyond that required for the refurbishment
activity will be scheduled concurrent with the repair activity.
Other compensatory actions that will be implemented include:
operator attention to the importance of protecting the operable
redundant train and support systems will be increased, selection of
beneficial Makeup Pump configurations, no elective maintenance will
be scheduled in the switchyard, and the establishment of fire
watches.
As described above in Item 1, a PSA has been performed to assess
the risk impact of an increase in Completion Time. Although the
proposed one-time change results in an increase in Core Damage
Frequency (CDF), and Large Early Release Frequency (LERF), the value
of these increases is considered as small (CDF) and very small
(LERF) in the current regulatory guidance.
Therefore, granting this LAR does not involve a significant
reduction in the margin of safety.
Based on the above, Progress Energy Florida, Inc. (PEF)
concludes that the proposed LAR presents a no significant hazards
consideration under the standards set forth in 10 CFR 50.92(c), and
accordingly, a finding of ``no significant hazards consideration''
is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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 T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Section Chief: Michael L. Marshall.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: October 29, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification 3.1.8, ``Scram Discharge Volume (SDV)
Vent and Drain Valves,'' to allow a vent or drain line with one
inoperable valve to be isolated instead of requiring the valve to be
restored to Operable status within 7 days.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on February 24, 2003 (68 FR 8637), on possible
amendments to revise the action for one or more SDV vent or drain lines
with an inoperable valve, 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 April 15,
2003 (68 FR 18294). The licensee affirmed the applicability of the
model NSHC determination in its application dated October 29, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
A change is proposed to allow the affected SDV vent and drain
line to be isolated when there are one or more SDV vent or drain
lines with one valve inoperable instead of requiring the valve to be
restored to operable status within 7 days. With one SDV vent or
drain valve inoperable in one or more lines, the isolation function
would be maintained since the redundant valve in the affected line
would perform its safety function of isolating the SDV. Following
the completion of the required action, the isolation function is
fulfilled since the associated line is isolated. The ability to vent
and drain the SDV is maintained and controlled through
administrative controls. This requirement assures the reactor
protection system is not adversely affected by the inoperable
valves. With the safety functions of the valves being maintained,
the probability or consequences of an accident previously evaluated
are not significantly increased.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Accident Previously
Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. Thus,
this change does not create the possibility of a new or different
kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change ensures that the safety functions of the SDV
vent and drain valves are fulfilled. The isolation function is
maintained by redundant valves and by the required action to isolate
the affected line. The ability to vent and drain the SDV is
maintained through administrative controls. In addition, the reactor
protection system will prevent filling of the SDV to the point that
it has insufficient volume to accept a full scram. Maintaining the
safety functions related to isolation of the SDV and insertion of
control rods ensures that the proposed change does not involve a
significant reduction in the margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Section Chief: M. Kotzalas (Acting).
[[Page 5248]]
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: December 27, 2004.
Description of amendment requests: The requested change will delete
Technical Specification (TS) 5.7.1.1.a, ``Occupational Radiation
Exposure Report,'' and TS 5.7.1.4, ``Monthly Operating Reports.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated December 27, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating letter
report of shutdown experience and operating statistics if the
equivalent data is submitted using an industry electronic database.
It also eliminates the TS reporting requirement for an annual
occupational radiation exposure report, which provides information
beyond that specified in NRC regulations. The proposed change
involves no changes to plant systems or accident analyses. As such,
the change is administrative in nature and does not affect
initiators of analyzed events or assumed mitigation of accidents or
transients. 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 a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve significance hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Robert A. Gramm.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: December 27, 2004.
Description of amendment requests: The proposed amendments would
revise the San Onofre Nuclear Generating Station (SONGS), Units 2 and 3
accident source term used in the design basis radiological consequences
analyses. These license amendments are requested in accordance with the
requirements of 10 CFR 50.67, which addresses the use of an Alternative
Source Term (AST) at operating reactors, and relevant guidance of
Regulatory Guide 1.183. These license amendments represent full-scope
implementation of the AST described in Regulatory Guide 1.183.
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 changes to the Facility Operating Licenses for San
Onofre Units 2 and 3 credit an Alternative Source Term (AST) for the
design basis radiological site boundary and control room dose
analyses. This change represents full scope implementation of the
AST as described in Regulatory Guide 1.183. The proposed changes to
the Facility Operating Licenses also expand the allowed use of fuel
failure estimates by Departure from Nucleate Boiling (DNB)
statistical convolution methodology from only the reactor coolant
pump sheared shaft event to the Updated Final Safety Analysis Report
(UFSAR) Chapter 15 non-Loss-of-Coolant-Accident (LOCA) events that
assume a loss of flow (i.e., a loss of AC power) and that fail fuel.
The proposed changes reflect the parameters used in the radiological
consequences calculations for the LOCA, Fuel Handling Accident
inside containment (FHA-IC), Fuel Handling Accident in the Fuel
Handling Building (FHA-FHB) and pre-trip Steam Line Break Outside
Containment (SLB-OC).
The purpose of this proposed change is to change the design
requirements for the Control Room Envelope (CRE). This proposed
change will allow an increase in the assumed amount of unfiltered
air inleakage through the CRE. Currently, design basis radiological
consequence analyses assume CRE inleakage of 0 cfm, plus an assumed
10 cubic feet per minute (cfm) inleakage due to ingress and egress
into the Control Room. Analyses to support this change demonstrate
acceptable post-accident dose consequences in the Control Room
assuming 990 cfm of CRE inleakage (plus 10 cfm due to ingress and
egress for a total of 1000 cfm).
This proposed change does not affect the precursors for
accidents or transients analyzed in Chapter 15 of the San Onofre
Units 2 and 3 UFSAR. Therefore, there is no increase in the
probability of accidents previously evaluated. The probability
remains the same because the accident analyses performed involve no
change to a system, component or structure that affects initiating
events for any UFSAR Chapter 15 accident evaluated.
A re-analysis of the UFSAR Chapter 15 LOCA, SLB-OC, FHA-IC, and
FHA-FHB events was conducted with respect to radiological
consequences. This re-analysis was performed in accordance with AST
methodology provided in Regulatory Guide (RG) 1.183 and with ARCON96
atmospheric dispersion methodology provided in RG 1.194. The
reanalysis consequences were expressed in terms of Total Effective
Dose Equivalent (TEDE) dose.
Implementation of the AST methodology, as described in 10 CFR
50.67, specifies control room, exclusion area boundary (EAB), and
low population zone (LPZ) dose acceptance criteria in terms of TEDE
dose. The dose acceptance criteria for specific events are specified
in RG 1.183. The revised analyses for all evaluated events meet the
applicable RG 1.183 TEDE dose acceptance criteria for AST
implementation.
The previous dose calculations analyzed the dose consequences to
thyroid and whole body as a result of postulated design basis
events. The previous control room dose calculations were shown to be
within the regulatory limits of 10 CFR 50 Appendix A General Design
Criterion 19 with respect to thyroid, beta-skin and whole body dose.
The previous LOCA and SLB offsite dose calculations were shown to be
within the regulatory limits of 10 CFR 100.11 with respect to
thyroid and whole body dose. The previous FHA-IC and FHA-FHB offsite
dose calculations were shown to be well within (i.e., less than 25
percent of) the regulatory limits of 10 CFR 100.11 with respect to
thyroid and whole body dose. RG 1.183 Footnote 7 provides a means to
compare the thyroid and whole body dose results of the previous
calculations with the TEDE results of the AST calculations. This
methodology requires multiplying the previous thyroid dose by 0.03
and adding the product to the previous whole body dose. The
resultant
[[Page 5249]]
``effective'' TEDE is then compared to the AST TEDE result. This
comparison is presented in Table 5-1.
The Table 5-1 comparison shows a decrease in dose consequences
when evaluated using AST methodology for all but the LOCA offsite
dose receptors. The LOCA EAB dose using AST methodology has
increased due to the requirement to calculate the maximum 2-hour
window EAB dose versus the previous requirement to calculate the 0
to 2 hour window EAB dose. The LOCA LPZ dose using AST methodology
has increased primarily due to changes in the AST Refueling Water
Storage Tank (RWST) iodine transport model. Although the LOCA EAB
and LPZ doses using AST methodology have increased, they remain
significantly below the 25 Rem TEDE offsite dose acceptance
criterion.
Table 5-1.--Comparison of Previous and AST Doses
------------------------------------------------------------------------
``Effective''
TEDE of previous
Event-dose receptor dose analyses AST TEDE (Rem)
(Rem)
------------------------------------------------------------------------
FHA-IC:
Control Room.................. 1.0 2.7 E-01
EAB........................... 2.0 8.0 E-01
LPZ........................... 5.6 E-02 2.3 E-02
FHA-FHB:
Control Room.................. 3.7 E-01 7.3 E-02
EAB........................... 6.6 E-01 2.1 E-01
LPZ........................... 1.9 E-02 6.1 E-03
LOCA:
Control Room.................. 4.5 2.7
EAB........................... 3.7 5.1
LPZ........................... 1.2 1.8
SLB-OC:
Control Room.................. (\1\) 2.1
EAB........................... 8.0 4.1
LPZ........................... (\1\) 0.1
------------------------------------------------------------------------
\1\ Not evaluated.
The proposed changes do not increase the probability of an
accident previously evaluated. The proposed changes result in dose
consequences that, if compared to previous ones, are in most cases
decreased and in other cases only slightly increased (using guidance
in footnote 7 of RG 1.183). However, the dose consequences of the
revised analyses are below the AST regulatory acceptance criteria.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The implementation of this proposed change does not create the
possibility of an accident of a different type than was previously
evaluated in the UFSAR. The proposed change credits the AST for the
design basis radiological site boundary and control room dose
analyses and expands the allowed use of fuel failure estimates by
DNB statistical convolution methodology from only the reactor
coolant pump sheared shaft event to the UFSAR Chapter 15 non-LOCA
events that assume a loss of flow (i.e., a loss of AC power) and
that fail fuel. The changes proposed do not change how Design Basis
Accident (DBA) events were postulated nor do the changes themselves
initiate a new kind of accident with a unique set of conditions. The
changes proposed are based on a re-analysis of offsite and control
room doses for four design basis accidents. The revised analyses are
consistent with the regulatory guidance established in RG 1.183. The
revised analyses utilize the most current understanding of source
term timing and chemical forms. Through this re-analysis, no new
accident initiator or failure mode was identified.
Therefore, this proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The implementation of this proposed amendment does not reduce
the margin of safety. The alternative source term radiological dose
consequence analyses utilize the regulatory acceptance criteria of
10 CFR 50 Appendix A General Design Criterion (GDC) 19 and 10 CFR
50.67, as specified in RG 1.183. These acceptance criteria have been
developed for the purpose of use in design basis accident analyses
such that meeting these limits demonstrates adequate protection of
public health and safety. An acceptable margin of safety is inherent
in these licensing limits. The radiological analyses results remain
within these regulatory acceptance criteria.
Therefore, there is no significant reduction in the margin of
safety as a result of the proposed amendment.
Based on the above, SCE concludes that the proposed amendments
present no significant hazards consideration under the standards set
forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no
significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: 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: November 12, 2004.
Description of amendment request: The proposed amendments would
revise Technical Specifications 3.1.7, ``Standby Liquid Control (SLC)
System,'' for Hatch Units 1 and 2. The proposed amendments would update
Figure 3.1.7-1 of Units 1 and 2 TS to reflect the increased
concentration of Boron-10 in the solution. Conforming revisions to
Bases B 3.1.7, ``Standby Liquid Control (SLC) System'' are also
included.
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
[[Page 5250]]
consequences of an accident previously evaluated?
This is a proposed change to Figure 3.1.7-1 of the Units 1 and 2
Technical Specifications. This figure is a graph of the weight
percent of Sodium Pentaborate solution in the Standby Liquid Control
(SLC) Tank, as a function of the gross volume of solution in the
tank. The figure is proposed to be changed in order to accommodate
an injection of Sodium Pentaborate solution into the reactor,
following an ATWS event, such that the concentration of Boron-10
atoms in the reactor will be 800 ppm natural Boron equivalent. This
is necessary to accommodate increased cycle energy requirements for
the Hatch Units 1 and 2 cores.
The proposed change to the Figure will not increase the
probability of an ATWS event because the curve has nothing to do
with the prevention of an ATWS event. The new requirements will
ensure that, in the future, the core will have adequate shutdown
margin to mitigate the consequences of an ATWS event.
Also, no systems or components designed to ensure the safe
shutdown of the reactor are being physically changed as a result of
this proposed TS change. In fact, no safety related systems or
components designed for the prevention of previously evaluated
events are being altered by the amendment.
As a result, the probability and consequences of an ATWS event,
or any other previously evaluated event, will not increase as a
result of this amendment.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
This proposed TS revision results in a change to the SLC TS
figure 3.7.1-1 requirements. However, this does not result in
physical changes to the SLC system. SLC pump operation, maintenance
and testing remain the same. Accordingly, no changes to the
operation, maintenance or surveillance procedures will result from
this TS revision request. Therefore, no new modes of operation are
introduced by this TS change.
Since no new modes of operation are introduced, the proposed
change does not create the possibility of a new or different type
event from any previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
This proposed TS change is being made to increase the boron
concentration requirements of the sodium pentaborate solution
injected into the reactor vessel following an Anticipated Transient
Without Scram (ATWS) event. The change is necessary due to new fuel
designs and higher energy requirements for fuel cycles. Therefore,
the change is being made to insure that shutdown requirements can be
met for the ATWS event. This will insure the margin of safety with
respect to ATWS will continue to be met.
Consequently, this proposed TS change will not result in a
decrease 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: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: John A. Nakoski.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: October 14, 2004.
Description of amendment request: The requested change will delete
Technical Specification (TS) 6.9.1.5 related to the annual
``Occupational Radiation Exposure Report,'' and TS 6.9.1.10, ``Monthly
Reactor Operating Report.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated October 14, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating letter
report of shutdown experience and operating statistics if the
equivalent data is submitted using an industry electronic database.
It also eliminates the TS reporting requirement for an annual
occupational radiation exposure report, which provides information
beyond that specified in NRC regulations. The proposed change
involves no changes to plant systems or accident analyses. As such,
the change is administrative in nature and does not affect
initiators of analyzed events or assumed mitigation of accidents or
transients. 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?
Response: No.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Michael L. Marshall, Jr.
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
[[Page 5251]]
Evaluation and/or Environmental Assessment as indicated. All of these
items are available for public inspection at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
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
PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by email to
[email protected].
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina; Carolina Power & Light Company, Docket No. 50-261, H.B.
Robinson Steam Electric Plant, Unit No. 2, Darlington County, South
Carolina
Date of application for amendments: December 19, 2003, as
supplemented January 14, 2004.
Brief description of amendments: The amendments allows entry into a
mode or other specified condition in the applicability of a technical
specification (TS), while in a condition statement and the associated
required actions of the TS, provided the licensee performs a risk
assessment and manages risk consistent with the program as proposed by
the industry's Technical Specification Task Force (TSTF) and is
designated TSTF-359.
Date of issuance: January 11, 2005.
Effective date: January 11, 2005.
Amendment Nos.: 233 and 260.
Facility Operating License Nos. DPR-71, DPR-62, and DPR-23.:
Amendments change the Technical Specifications.
Date of initial notice in Federal Register: The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated January 11, 2005.
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: October 26, 2004, as
supplemented on December 22, 2004.
Brief description of amendment: The amendment revises Technical
Specification 3.7.11, ``Control Room Ventilation System (CRVS),'' to
allow, on a one-time basis, an extension of the allowed outage time to
support placement of the CRVS in an alternate configuration for tracer
gas testing. The proposed amendment would also allow self-contained
breathing apparatus and potassium iodide pills to be used as
compensatory measures for the control room operators in the event that
the tracer gas test results are not bounded by the dose consequence
evaluations.
Date of issuance: January 19, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 223.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 8, 2004 (69 FR
64792).
The December 22 letter provided 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 as published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 19, 2005.
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: November 25, 2003.
Brief description of amendments: The amendments modify the Limerick
Generating Station, (LGS) Units 1 and 2, Technical Specifications (TSs)
contained in Appendix A to Operating License Nos. NPF-39 and NPF-85,
respectively. The amendments add a footnote to the LGS TS 3.4.3.2.e to
indicate that reactor coolant system (RCS) pressure isolation valve
leakage is excluded from any other allowable RCS operational leakage
specified in LGS TS 3.4.3.2.
Date of issuance: January 18, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 172 and 134.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the TSs.
Date of initial notice in Federal Register: February 3, 2004 (69 FR
5203).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 18, 2005.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: March 31, 2004.
Brief description of amendment: This amendment revised Technical
Specification (TS) requirements for mode change limitations in Limiting
Condition for Operation 3.0.4 and Surveillance Requirement 3.0.4 to
adopt the provisions of Industry TS Task Force (TSTF) change TSTF-359,
``Increase Flexibility in Mode Restraints.''
Date of issuance: January 6, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 131.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 6, 2004 (69 FR
40675).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 6, 2005.
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: December 19, 2003.
Brief description of amendment: The amendment modifies TS
requirements to adopt the provisions of Industry/TS Task Force (TSTF)
change TSTF-359, ``Increased Flexibility in Mode Restraints.'' The
availability of TSTF-359 for adoption by licensees was announced in the
Federal Register on April 4, 2003 (68 FR 16579).
Date of issuance: January 11, 2005.
Effective date: As of the date of issuance and shall be implemented
within 180 days of issuance.
Amendment No.: 215.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 17, 2004 (69
FR 7523).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 11, 2005.
No significant hazards consideration comments received: No.
[[Page 5252]]
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of application for amendments: April 23, 2004.
Brief description of amendments: The amendments revise several
Technical Specification (TS) Allowed Outage Times for TS 3.3.3,
Accident Monitoring Instrumentation, to be consistent with the
Completion Times in the related Specification in NUREG-1431, Revision
2, ``Standard Technical Specifications Westinghouse Plants (the
Improved Standard Technical Specifications, or ISTS).''
Date of issuance: January 6, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos: 227 and 223.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: May 25, 2004 (69 FR
29767).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 6, 2005.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of application for amendment: December 23, 2003.
Brief description of amendment: The amendment revises Technical
Specification (TS) requirements to adopt the provisions of the TS Task
Force (TSTF) change TSTF-359, regarding increased flexibility in mode
changes. The availability of TSTF-359 for adoption by licensees was
announced in the Federal Register on April 4, 2003 (68 FR 16579).
Date of issuance: January 10, 2005.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 255.
Facility Operating License No. DPR-49: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 16, 2004 (69
FR 55844).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 10, 2005.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: September 30, 2004.
Brief description of amendments: The amendments delete Technical
Specification (TS) 6.9.1.2, ``Occupational Radiation Exposure Report,''
and TS 6.9.1.5, ``Monthly Operating Reports,'' as described in the
Notice of Availability published in the Federal Register on June 23,
2004 (69 FR 35067).
Date of issuance: January 5, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: Unit 1-168; Unit 2-157.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: October 26, 2004 (69 FR
62478).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 5, 2005.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: February 3, 2004 as supplemented by
letter dated December 1, 2004.
Brief description of amendments: The amendments modify Technical
Specifications (TSs) requirements to adopt the provisions of Industry/
TS Task Force (TSTF) change TSTF-359, ``Increase Flexibility in Mode
Restraints.'' The availability of TSTF-359 for adoption by licensees
was announced in the Federal Register on April 4, 2003 (68 FR 16579).
Date of issuance: January 10, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: Unit 1-170; Unit 2-158.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: March 2, 2004 (69 FR
9865).
The supplement dated December 1, 2004, 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 as
published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 10, 2005.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an
[[Page 5253]]
opportunity to provide for public comment on its no significant hazards
consideration determination. In such case, the license amendment has
been issued without opportunity for comment. If there has been some
time for public comment but less than 30 days, the Commission may
provide an opportunity for public comment. If comments have been
requested, it is so stated. In either event, the State has been
consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room (PDR), located at One White Flint North, Public File Area 01F21,
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly
available records will be accessible from the Agencywide Documents
Access and Management System's (ADAMS) Public Electronic Reading Room
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the PDR Reference
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, the licensee may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene. Requests for a hearing and a petition for leave
to intervene shall be filed in accordance with the Commission's ``Rules
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2.
Interested persons should consult a current copy of 10 CFR 2.309, which
is available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a
request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, 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 general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
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. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ 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/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
---------------------------------------------------------------------------
\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the
[[Page 5254]]
authority to act for the petitioners/requestors with respect to that
contention.
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. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) e-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. 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 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 requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer or the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).
STP Nuclear Operating Company, Docket No. 50-498, South Texas Project,
Unit 1, Matagorda County, Texas
Date of amendment request: January 6, 2005.
Description of amendment request: The amendment revises Technical
Specification (TS) 3.7.4, ``Essential Cooling Water System,'' and the
associated TS for systems supported by the Essential Cooling Water
(ECW), to extend the allowed outage time for an additional 7 days for
ECW Train B as a one-time change for the purpose of making repairs to
the Train B ECW pump.
Date of issuance: January 10, 2005.
Effective date: Effective as of the date of issuance and shall be
implemented immediately.
Amendment No.: 169.
Facility Operating License No. NPF-76: Amendment revises the
technical specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No.
The Commission's related evaluation of the amendment, finding of
emergency circumstances, state consultation, and final NSHC
determination are contained in a safety evaluation dated January 10,
2005.
Attorney for licensee: A.H. Gutterman, Morgan, Lewis & Bockius,
1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: Michael K. Webb, Acting.
Dated at Rockville, Maryland, this 24th day of January 2005.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management Office of Nuclear
Reactor Regulation.
[FR Doc. 05-1574 Filed 1-31-05; 8:45 am]
BILLING CODE 7590-01-P