[Federal Register Volume 70, Number 30 (Tuesday, February 15, 2005)]
[Notices]
[Pages 7762-7777]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-2788]
-----------------------------------------------------------------------
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 20, 2005, through February 3, 2005.
The last biweekly notice was published on February 1, 2005 (70 FR
5233).
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 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
[[Page 7763]]
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 transmitted either by
means of facsimile transmission to (301) 415-3725 or by email 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 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; Docket No. 50-400, Shearon Harris Nuclear Power Plant, Unit
1, Wake and Chatham Counties, 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 amendments request: November 17, 2004.
Description of amendments request: The requested change would
delete Technical Specification (TS) 5.6.1, ``Occupational Radiation
Exposure Report,'' and TS 5.6.4, ``Monthly Operating Reports,'' for the
Brunswick and H. B. Robinson plants. The equivalent change is being
requested for the Shearon Harris facility by deleting TS 6.9.1.2.a and
TS 6.9.1.2.c under ``Annual Reports'' and TS 6.9.1.5, ``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 November 17, 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 (TS)
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
[[Page 7764]]
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 a 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.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: January 5, 2005.
Description of amendment request: The proposed amendments would
revise the Technical Specification (TS) 5.5.19 associated with the Lee
Combustion Turbine (LCT) testing program. TS 5.5.19.b currently
requires verification that an LCT can supply the equivalent of one
Unit's maximum safeguard loads, plus two Units' Mode 3 loads, when
connected to the system grid every 12 months. In the proposed
amendments, this requirement would be more clearly specified as ``plus
two Units' safe shutdown loads.''
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:
Duke proposes to revise TS 5.5.19.b to clarify the Lee
Combustion Turbine (LCT) testing requirements. The proposed change
makes the wording of the test requirement consistent with the UFSAR
[Updated Final Safety Analysis Report] and the original wording of
the TS requirement before administrative changes were made in
Amendment 232, 232, 231, and Amendment 300, 300, and 300. LCT
testing has no impact on the probability of an accident analyzed in
the UFSAR. The LCT can be credited to mitigate the consequences of
an accident analyzed in the UFSAR. However, this clarification of
LCT testing requirements has no impact on its ability to mitigate
the consequences of an accident. As such, the proposed LAR [license
amendment request] does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any kind of accident previously evaluated:
Duke proposes the revise TS 5.5.19.b to clarify the Lee
Combustion Turbine (LCT) testing requirements. The proposed change
makes wording of the test requirement consistent with the UFSAR and
the original wording of the TS requirement before administrative
changes were made in Amendment 232, 232, 231, and changes were made
in Amendment 300, 300, and 300. These changes do not alter the
nature of events postulated in the Safety Analysis Report nor do
they introduce any unique precursor mechanisms. Therefore, the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
(3) Involve a significant reduction in a margin of safety.
The proposed TS change does not unfavorably affect any plant
safety limits, set points, or design parameters. The changes also do
not unfavorably affect the fuel, fuel cladding, RCS [reactor coolant
system], or containment integrity. Therefore, the proposed TS
change, which clarifies TS requirements associated with the LCT
testing program, does not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anne W. Cottingham, Winston and Strawn LPP,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: John A. Nakoski.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: July 29, 2004.
Description of amendment request: The proposed amendment would
delete the requirements from the technical specifications (TS) to
maintain a hydrogen dilution system, a hydrogen purge system, 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 TS 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 related vent and purge systems 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
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 no significant hazards
consideration determination in its application dated July 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
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 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
[[Page 7765]]
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 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.
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, 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 monitor requirements,
including removal of these requirements from the 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 [dilution/purge
system for Davis Besse] 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 [dilution/purge system for Davis Besse] and
hydrogen monitor equipment was intended to mitigate a design-basis
hydrogen release. The hydrogen recombiner [dilution/purge system for
Davis Besse] 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 [dilution/purge
system for Davis Besse] 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 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.
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: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Gene Y. Suh.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: December 20, 2004.
Description of amendment request: The proposed change would revise
Technical Specification (TS) 3/4.9.2, ``Refueling Operations--
Instrumentation,'' concerning source range neutron flux monitors to be
consistent with Improved Standard Technical Specifications.
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 changes affect the Limiting Condition for Operation [LCO]
for Refueling Operations--Instrumentation, in particular, the LCO
sections pertaining to the source range neutron flux detectors will
be changed to be more like the corresponding sections in the
Improved Standard Technical Specifications. The source range neutron
flux detectors have no control functions and are therefore not
accident initiators. Consequently, the proposed changes will have no
impact on the probability of any accident previously evaluated. The
detectors are not credited in mitigating the consequences of any
accident; therefore, the proposed changes will have no impact on the
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 changes affect the Limiting Condition for Operation for
Refueling Operations--Instrumentation, in particular, the source
range neutron flux detectors. The source range neutron flux
detectors will continue to operate in the same manner as previously
considered. Accident initial conditions and assumptions remain as
previously analyzed.
The proposed changes do not introduce any new or different
accident initiators. Therefore, the proposed changes do not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The changes affect the Limiting Condition for Operation for
Refueling Operations--Instrumentation; in particular, the source
range neutron detectors. These detectors have no control functions,
and are not credited in mitigating the consequences of any accident.
The source range neutron detectors are not associated with a safety
limit. In addition, the proposed changes to TS will not result in
design changes to the source range neutron detectors or in changes
to how the source range detectors are used. Therefore, the proposed
changes will not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Gene Y. Suh.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: January 5, 2005.
Description of amendment request: The license amendment would
revise Technical Specification 3/4.3.2.1, ``Safety Features Actuation
System [SFAS] Instrumentation,'' to permit a single inoperable SFAS
functional unit to be placed in a bypassed condition indefinitely.
[[Page 7766]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change would permit a single SFAS instrument string
functional unit to be placed in bypass indefinitely. The primary
function of SFAS is to monitor station conditions and actuate the
engineered safety features when needed in order to prevent or limit
fission product and energy release from the core, to isolate the
containment vessel, and to initiate the operation of the Engineered
Safety Features (ESF) equipment in the event of a loss-of-coolant
accident (LOCA).
The SFAS is a possible accident initiator in that an inadvertent
system level actuation could result in a transient or accident. The
existing Technical Specification requirements for SFAS allow
operation indefinitely with a single SFAS functional unit in trip,
which results in a 1-out-of-3 channel logic. In this condition, the
spurious actuation in one of the three remaining corresponding
functional unit would result in an inadvertent system level
actuation. Under the proposed change, indefinite operation in a 2-
out-of-3, 1-out-of-3, or 1-out-of-2 channel logic would be allowed.
The likelihood of a spurious system level actuation for any of the
configurations allowed under the proposed change is no greater than
the likelihood of spurious actuation under the 1-out-of-3 channel
logic allowed under the existing Technical Specification
requirements. Therefore, operation of the SFAS actuation from that
permitted by the existing Technical Specifications.
Under the proposed change, the SFAS will continue to perform
this function with a high level of reliability. The proposed change
would allow operation of the SFAS in a condition with reduced
redundancy from what is currently required by the Technical
Specifications. Operation of the SFAS with reduced redundancy was
evaluated against the design criteria to which the system was
designed. The design criteria applicable to the SFAS, including the
single failure criterion, continue to be met. The proposed change
does not prevent the SFAS from mitigating the consequences of
previously analyzed accidents.
The proposed change would not increase the likelihood of an
inadvertent SFAS actuation. The proposed change would not prevent
the SFAS from mitigating the consequences of previously analyzed
accidents. 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 alter the SFAS design function or
the manner in which that function is performed. Under the proposed
change, the SFAS will continue to perform its function with a high
degree of reliability. No new failure modes or accident initiators
are created by the proposed change. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change would allow operation of the SFAS in a
condition with reduced redundancy from what is currently required by
the Technical Specifications. Operation of the SFAS with reduced
redundancy was evaluated against the design criteria to which the
system was designed. This evaluation shows that with the SFAS in the
conditions permitted by the proposed change, the SFAS still
satisfies all the applicable design criteria, including the single
failure criterion. 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: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Gene Y. Suh.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: January 11, 2005.
Description of amendment request: The proposed amendment would
revise the Updated Safety Analysis Report (USAR) by modifying the
design requirements for protection from tornado missiles. Specifically,
the proposed amendment would allow certain structures, systems, and
components that are not currently provided with physical protection
from tornado-induced missiles to be evaluated for acceptability based
on the Electrical Power Research Institute ``Tornado Missile Risk
Evaluation Methodology'' (TORMIS).
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 amendment would reflect use of the Electric Power
Research Institute (EPRI) Topical Report ``Tornado Missile Risk
Evaluation Methodology'' (EPRI NP-2005), Volumes I and II. As noted
in the NRC Safety Evaluation on this report dated October 26, 1983,
``The current licensing criteria governing tornado missile
protection are contained in Standard Review Plan (SRP) Sections
3.5.1.4 and 3.5.2. These criteria generally specify that safety-
related systems be provided positive tornado missile protection
(barriers) from the maximum credible tornado threat. However, SRP
Section 3.5.1.4 includes acceptance criteria permitting relaxation
of the above deterministic guidance, if it can be demonstrated that
the probability of damage to unprotected essential safety-related
features is sufficiently small.''
``Certain Operating License (OL) applicants and operating
reactor licensees have chosen to demonstrate compliance with tornado
missile protection criteria for certain portions of the plant * * *
by providing a probabilistic analysis which is intended to show a
sufficiently low risk associated with tornado missiles. Some* * *
have utilized the tornado missile probabilistic risk assessment
(PRA) methodology developed by'' EPRI in the Topical Report listed
above. The NRC noted that this report ``can be utilized when
assessing the need for positive tornado missile protection for
specific safety-related plant features.'' This methodology has
subsequently been utilized in nuclear power plant licensing actions.
As permitted in NRC Standard Review Plan (NUREG-0800) sections,
the total probability will be maintained below an allowable level,
i.e., an acceptance criteria threshold, which reflects an extremely
low probability of occurrence. The DBNPS [Davis-Besse Nuclear Power
Station] approach assumes that if the probability calculation result
for the total plant identifies that the cumulative probability of
tornado missiles striking an unprotected portion of a safety system
or component required for safe shutdown in the event of a tornado
exceeds 10-6 per year, then unique missile barriers would
need to be installed to lower the total probability below the
acceptance criteria of 10-6 per year.
With respect to the probability of occurrence of an accident
previously evaluated in the USAR, the possibility of a tornado
reaching the DBNPS site and causing damage to plant structures,
systems, and components is an event considered in the USAR. The
changes being proposed herein do not affect the probability that the
natural phenomena (a tornado) will reach the plant, but they do,
from a licensing basis perspective, affect the probability that
missiles generated by the winds of the tornado might strike certain
plant systems or components. As recently determined, there are a
limited number of safety-related components that could theoretically
be struck by a tornado generated missile. The
[[Page 7767]]
total (cumulative) probability of a tornado missile striking an
unprotected component will be maintained below an extremely low
acceptance criteria to ensure overall plant safety. Due to the
extremely low probability of a tornado missile impacting an
essential component, the small increase in the probability of
accident initiation is not considered significant.
With respect to the consequences of an accident previously
evaluated, there is an extremely low probability of a malfunction of
an unprotected essential component due to tornado missile impact.
Due to (1) the extremely low probability of a tornado missile
striking essential equipment as calculated by TORMIS, and (2) the
low probability that any tornado missile strikes would cause
sufficient damage to prevent essential equipment from performing its
accident-mitigating function, a loss of accident mitigation
capability is not considered credible. Therefore, the radiological
consequences of accidents are not significantly affected.
The proposed change is not considered to constitute a
significant increase in the probability of occurrence or the
consequences of an accident, due to the extremely low total
probability of a tornado missile strike and thus an extremely low
probability of a radiological release. Therefore, the proposed
change does not involve a significant increase in the probability or
consequences of previously evaluated 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 possibility of a tornado reaching the DBNPS site is a design
basis event considered in the USAR. This change involves recognition
of the acceptability of performing tornado missile probability
calculations in accordance with established regulatory guidance. The
change therefore deals with an established design basis event (the
tornado). Therefore, the proposed change would not contribute to the
possibility of a new or different kind of accident from those
previously analyzed.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This request does not involve a significant reduction in a
margin of safety. The existing licensing basis for the DBNPS with
respect to the design basis event of a tornado reaching the plant is
to provide positive missile barriers for all systems and components
required for safe shutdown in the event of a tornado. With the
change, it will be recognized that there is an extremely low
probability, below an established acceptance limit, that a limited
subset of these systems and components could be struck. The change
to missile protection based on extremely low probability (less than
1 x 10-6 per year cumulative strike probability) of
occurrence of tornado generated missile strikes on portions of these
systems and components is not considered to constitute a significant
decrease in the margin of safety due to that extremely low
probability. Therefore, the changes associated with this license
amendment do not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Gene Y. Suh.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: September 10, 2004.
Description of amendment request: The proposed amendment would
delete the requirements from the technical specifications (TS) 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 TS 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 related vent and purge systems 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 no significant hazards consideration determination in its
application dated September 10, 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, 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 monitor requirements,
including removal of these requirements from the TS, does not
involve a significant increase in the probability or the
consequences of any accident previously evaluated.
[[Page 7768]]
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: David W. Jenkins, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Gene Y. Suh.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: November 17, 2004.
Description of amendment request: The requested change would delete
Technical Specification (TS) 5.7.1.1.a, ``Occupational Radiation
Exposure Report,'' and TS 5.7.1.2, ``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 November 17, 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 (TS)
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 a 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.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: August 6, 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 August 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:
[[Page 7769]]
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 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 [a] 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: Daniel F. Stenger, Ballard Spahr Andrews &
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC
20005.
NRC Section Chief: Richard J. Laufer.
Nine Mile Point Nuclear Station, LLC, Docket Nos. 50-220 and 50-410,
Nine Mile Point Nuclear Station, Unit Nos. 1 and 2 (NMP1 and NMP2),
Oswego County, New York
Date of amendment request: January 24, 2005.
Description of amendment request: The licensee proposed amendments
to delete Sections 6.6.1 and 5.6.1, ``Occupational Radiation Exposure
Report,'' and Sections 6.6.4 and 5.6.4, ``Monthly Operating Reports,''
from the NMP1 and NMP2 Technical Specifications, respectively. 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 January 24, 2005.
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 by referencing the model NSHC analysis published by the
NRC staff. The model NSHC analysis is reproduced 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.
[[Page 7770]]
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 has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Richard J. Laufer.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: October 28, 2004.
Brief description of amendments: The proposed amendment deletes the
requirements from the Technical Specifications (TS) 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 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
model NSHC determination in its application dated October 28, 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 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.
[[Page 7771]]
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Michael K. Webb (Acting).
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: November 4, 2004.
Description of amendment request: The proposed changes would
relocate the inservice testing requirements, remove the inservice
inspection requirements, and add a Bases Control Program to the
Administrative Controls section of the Technical Specifications (TS).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of Surry Units 1 and 2 in accordance with the
proposed Technical Specifications change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed change is administrative in nature, and station
operations are not being affected. The ASME [American Society of
Mechanical Engineers] Code requirements are established, reviewed
and approved by ASME, the industry and ultimately endorsed by the
NRC for inclusion into 10 CFR 50.55a. Updates to the ASME Code
reflect advances in technology and consider information obtained
from plant operating experience to provide enhanced inspection and
testing. Thus, the proposed change only modifies TS to appropriately
reference the recently NRC approved Inservice Testing Program for
the fourth interval at Surry Power Station. Consequently, the
probability or consequences of an accident previously evaluated are
not increased.
2. The proposed Technical Specifications change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
As noted above, the proposed change is administrative in nature,
and no new accident precursors are being introduced. Since the
inservice testing will continue to be performed in accordance with
an NRC approved program, adequate assurance is provided to ensure
the safety-related pumps and valves would operate as required. No
new testing is required that could create a new or different type of
accident. Consequently, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed Technical Specifications change does not involve
a significant reduction in a margin of safety.
Performing inservice testing of pumps and valves to the NRC
approved program for the fourth interval at Surry Power Station
provides adequate assurance that the safety-related pumps and valves
will continue to perform their intended safety function. This is an
administrative change in nature and as such does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Section Chief: John A. Nakoski.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant
(WBN), Unit 1, Rhea County, Tennessee
Date of amendment request: September 15, 2004.
Brief description of amendment request: In accordance with
Technical Specification Task Force (TSTF) 285, Charging Pump Swap Low-
Temperature Over-Pressurization Allowance, LCO 3.4.12, Cold
Overpressure Mitigation System (COMS), is being revised to modify and
relocate two notes in the WBN Technical Specifications. The changes are
all administrative, except a change which would allow two charging
pumps to be made capable of injecting into the Reactor Coolant System
to support pump swap operations for a period not to exceed one hour
instead of the currently allowed 15 minutes.
Date of publication of individual notice in Federal Register:
February 1, 2005 (70 FR 5226).
Expiration date of individual notice: March 3, 2005 (public
comments) and April 4, 2005 (hearing requests).
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (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,
[[Page 7772]]
(301) 415-4737 or by e-mail to [email protected].
Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-318, Calvert
Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland
Date of application of amendment: September 30, 2003.
Description of amendment request: The amendment modifies Technical
Specification (TS) 4.3.1, ``Criticality,'' adds TS 3.7.16, ``Spent Fuel
Pool Boron Concentration,'' and adds TS 3.7.17, ``Spent Fuel Pool
Storage.'' Specifically, the amendment increases the maximum enrichment
limit of the fuel assemblies that can be stored in the Unit 2 spent
fuel pool by taking credit for soluble boron, burnup, and configuration
control in maintaining acceptable margins of subcriticality.
Date of issuance: January 27, 2005.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 246.
Renewed License No. DPR-69: Amendment revised the Technical
Specifications.
Date of initial notice in Federal Register: January 20, 2004 (69 FR
2739).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 27, 2005.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: July 26, 2004, as supplemented
January 26, 2005.
Brief description of amendments: These amendments revise the
Technical Specifications by eliminating the requirements associated
with hydrogen and oxygen monitors.
Date of issuance: February 2, 2005.
Effective date: As of its date of issuance, and shall be
implemented within 120 days.
Amendment Nos.: 234 and 261.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: August 31, 2004 (69 FR
53100). The January 26, 2005, supplement contained clarifying
information only and did not change the initial proposed no significant
hazards consideration determination or expand the scope of the initial
application.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 2, 2005.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: February 25, 2003, as
supplemented June 9, and July 30, 2003, and September 13, 2004.
Brief description of amendments: The amendments revised the
Technical Specifications to incorporate a Steam Generator (SG) program
that defines a performance-based approach to maintaining SG tube
integrity. The SG program includes performance criteria that define the
basis for tube integrity and provides reasonable assurance that SG
tubing will remain capable of fulfilling its safety function of
maintaining reactor coolant system pressure boundary integrity. The
proposed amendments add a new TS for SG tube integrity (3.4.18) and
revise the TS for reactor coolant operation leakage (3.4.13), SG tube
surveillance program (5.5.9), and SG tube inspection report (5.6.8).
Date of issuance: January 13, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 218 and 212.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: July 8, 2003 (68 FR
40712).
The supplements dated June 9, and July 30, 2003, and September 13,
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 13, 2005.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: August 18, 2004.
Brief description of amendments: The amendments revised the
Technical Specifications to remove references to Safety Injection Steam
Line Pressure-Low.
Date of issuance:
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 224 and 206.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: November 9, 2004 (69
FR 64987).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 27, 2005.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Units 1 and 2, Ogle County, Illinois; Docket Nos. STN
50-456 and STN 50-457, Braidwood Station, Units 1 and 2, Will County,
Illinois
Date of application for amendments: August 15, 2003, as
supplemented on April 9, 2004.
Brief description of amendments: The amendments revise Technical
Specification (TS) 3.4.15, ``RCS Leakage Detection Instrumentation'',
to require one containment sump monitor and one containment atmosphere
particulate radioactivity monitor to be operable in Modes 1, 2, 3, and
4.
Date of issuance: January 14, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 140, 133.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 28, 2003 (68
FR 61477).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 14, 2005.
No significant hazards consideration comments received: No.
Power and Light Company, et al., Docket No. 50-389, St. Lucie Plant,
Unit No. 2, St. Lucie County, Florida
Date of application for amendment: December 2, 2003, as
supplemented by letters dated September 14 and December 10, 2004, and
January 7, 2005.
Brief description of amendment: This amendment revised the
Technical Specifications (TSs) to permit operation
[[Page 7773]]
with a reduced reactor coolant system flow corresponding to a steam
generator (SG) tube plugging level of 30-percent per SG. This amendment
also includes the transition to Westinghouse Reload Safety Evaluation
Methodology (WCAP-9272).
Date of issuance: January 31, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 138.
Renewed Facility Operating License No. NPF-16: Amendment revised
the TSs.
Date of initial notice in Federal Register: March 18, 2004 (69 FR
12873).
The September 14 and December 10, 2004, and January 7, 2005,
supplements did not affect the original proposed no significant hazards
determination, or expand the scope of the request as noticed in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 31, 2005.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station, Unit No. 1, Oswego County, New York
Date of application for amendment: April 19, 2004, as supplemented
on July 16, 2004.
Brief description of amendment: The amendment revised Section 3/
4.6.2, ``Protective Instrumentation,'' to establish a 24-month
operating cycle calibration frequency for the intermediate range
monitor instrumentation. In addition, the amendment authorized
relocation of the limiting conditions for operation and surveillance
requirements for certain control rod withdrawal block instruments from
Section 3/4.6.2 to the Updated Final Safety Analysis Report.
Date of issuance: January 25, 2005.
Effective date: January 25, 2005.
Amendment No.: 186.
Facility Operating License No. DPR-63: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 25, 2004 (69 FR
29769).
The July 16, 2004, letter provided clarifying information within
the scope of the original application and did not change the staff's
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated January 25, 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: October 5, 2004.
Brief description of amendment: The amendment deletes Technical
Specification (TS) 5.6.1, ``Occupational Radiation Exposure Report,''
and TS 5.6.4 ``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 31, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 256.
Facility Operating License No. DPR-49: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 9, 2004 (69
FR 64989).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 31, 2005.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: January 30, 2004.
Brief description of amendment: The amendment changes the Technical
Specifications (TSs) to (1) clarify the permissive setpoint for the
source range monitor detector-not-fully-inserted rod block bypass, (2)
correct a typographical error in the surveillance requirement for
suppression pool temperature monitoring, (3) clarify the setpoint for
the pressure suppression chamber-reactor building vacuum breakers
instrumentation, (4) clarify the operating force requirements for the
pressure suppression chamber-drywell vacuum breakers surveillance test,
and (5) make corrections resulting from license Amendments 130 and 132.
Date of issuance: January 28, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 141.
Facility Operating License No. DPR-22. Amendment revised the TSs.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19573).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 28, 2005.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: October 5, 2004.
Brief description of amendment: The amendment deletes technical
specification (TS) 6.7.A.2, ``Requirement to submit an Occupational
Radiation Exposure Report,'' TS 6.7.A.3, ``Requirement to submit a
Monthly Operating Report,'' and TS 6.7.A.6, ``Requirement to report
safety/relief valve failures and challenges'' as described in the
Notice of Availability published in the Federal Register on June 23,
2004 (69 FR 35067).
Date of issuance: February 1, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 142.
Facility Operating License No. DPR-22. Amendment revised the TSs.
Date of initial notice in Federal Register: November 9, 2004 (69 FR
64989).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 1, 2005.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of application for amendment: October 5, 2004.
Brief description of amendment: The amendment deletes technical
specification 5.6.1, ``Occupational Radiation Exposure Report,'' and TS
5.6.4 ``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 10, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 220.
Facility Operating License No. DPR-20: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 9, 2004 (69 FR
64989).
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.
[[Page 7774]]
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of application for amendment: January 30, 2004.
Brief description of amendment: The amendment eliminates
requirements for hydrogen recombiners and relocates the requirements
for hydrogen monitors to the licensee's Commitment Management Program.
Date of issuance: January 11, 2005.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 221.
Facility Operating License No. DPR-20: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 2, 2004 (69 FR
9862).
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.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: October 5, 2004.
Brief description of amendments: The amendments delete technical
specification (TS) 5.6.1, ``Occupational Radiation Exposure Report,''
and TS 5.6.4 ``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 31, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 168, 158.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 9, 2004 (69 FR
64989).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 31, 2005.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: February 13, 2004, as
supplemented by letters dated November 5 and December 10, 2004.
Brief description of amendments: The amendments revise Technical
Specifications (TSs) 3.3.1, ``Reactor Trip System (RTS)
Instrumentation,'' 3.3.2, ``Engineered Safety Feature Actuation System
(ESFAS) Instrumentation,'' and 3.3.6, ``Containment Ventilation
Isolation Instrumentation,'' to adopt the completion time, test bypass
time, and surveillance frequency time changes approved by the NRC in
Topical Reports WCAP-14333-P-A, ``Probabilistic Risk Analysis of the
RPS [reactor protection system] and ESFAS Test Times and Completion
Times,'' and WCAP-15376-P-A, ``Risk-Informed Assessment of the RTS and
ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and
Completion Times.'' The amendments revise the required actions for
certain action conditions; increase the completion times for several
required actions (including some notes); delete notes in certain
required actions; and increase frequency time intervals (including
certain notes) in several surveillance requirements.
Date of issuance: January 31, 2005.
Effective date: January 31, 2005, and shall be implemented within
180 days of the date of issuance.
Amendment Nos.: Unit 1--179; Unit 2--181.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 30, 2004 (69 FR
16622). The supplemental letters dated November 5 and December 10,
2004, provided clarifying information that did not change the scope of
the original application as noticed or the NRC staff's original
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated January 31, 2005.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch Nuclear
Plant, Unit 2, Appling County, Georgia
Date of application for amendments: April 26, 2004, as supplemented
by letters dated August 17 and September 7, 2004.
Brief description of amendments: The amendment revised the
Technical Specification Section 5.5.12, ``Primary Containment Leakage
Rate Testing Program'' to reflect a one-time deferral of the Type A
Containment Integrated Leak Rate Test (ILRT). This change extends the
10-year interval between ILRTs to 15 years from the previous ILRT.
Date of issuance: February 1, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 187.
Renewed Facility Operating License No. NPF-5: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 3, 2004 (69 FR
46591).
The supplements dated August 17 and September 7, 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 February 1, 2005.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: December 17, 2003, as
supplemented by letters dated October 28 and November 16, 2004.
Brief description of amendment: The amendment revises TSs 3.3.1,
``Reactor Trip System (RTS) Instrumentation,'' 3.3.2, ``Engineered
Safety Feature Actuation System (ESFAS) Instrumentation,'' and 3.3.9,
``Boron Dilution Mitigation System (BDMS)'' to adopt the completion
time, test bypass time, and surveillance time interval changes in NRC-
approved WCAP-14333-P-A, ``Probabilistic Risk Analysis of the RPS
[reactor protection system] and ESFAS Test Times and Completion
Times,'' and WCAP-15376-P-A, ``Risk-Informed Assessment of the RTS and
ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and
Completion Times.'' The TS changes revise required actions for certain
action conditions; increase the completion times for several required
actions (including some notes); delete notes in certain required
actions; increase frequency time intervals (including certain notes) in
several surveillance requirements (SRs); add an action condition and
required actions; revise notes in certain SRs; and revise Table 3.3.2-
1. There is also an administrative correction to the format of the TSs.
[[Page 7775]]
Date of issuance: January 31, 2005.
Effective date: January 31, 2005, and shall be implemented within
120 days of its date of issuance.
Amendment No.: 165.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 3, 2004 (69 FR
5211).
The supplemental letters dated October 28 and November 16, 2004,
provided clarifying information that did not change the scope of the
original application as noticed or the NRC staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 31, 2005.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: December 15, 2003, as supplemented by
letters dated October 7 and November 12, 2004.
Brief description of amendment: The amendment revises Technical
Specifications (TSs) 3.3.1, ``Reactor Trip System (RTS)
Instrumentation,'' and 3.3.2, ``Engineered Safety Feature Actuation
System (ESFAS) Instrumentation,'' to adopt the completion time, test
bypass time, and surveillance frequency time changes approved by the
NRC in Topical Reports WCAP-14333-P-A, ``Probabilistic Risk Analysis of
the RPS [reactor protection system] and ESFAS Test Times and Completion
Times,'' and WCAP-15376-P-A, ``Risk-Informed Assessment of the RTS and
ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and
Completion Times.'' The amendment revises the required actions for
certain action conditions; increase the completion times for several
required actions (including some notes); delete notes in certain
required actions; and increase frequency time intervals (including
certain notes) in several surveillance requirements.
Date of issuance: January 31, 2005.
Effective date: January 31, 2005, and shall be implemented within
180 days of the date of issuance.
Amendment No.: 156.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 3, 2004 (69 FR
5212).
The supplemental letters dated October 7 and November 12, 2004,
provided clarifying information that did not change the scope of the
original application as noticed or the NRC staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 31, 2005.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: October 7, 2004.
Brief description of amendment: The amendment revises Section 5.3,
``Unit Staff Qualifications,'' of the technical specifications (TSs) to
add the qualification requirements for the shift manager and the
control room supervisor. In addition, based on a comparison review
performed by the NRC and Wolf Creek Nuclear Operating Corporation
personnel, editorial corrections are being made to the TSs.
Date of issuance: January 31, 2005.
Effective date: January 31, 2005, and shall be implemented within
90 days from the date of issuance.
Amendment No.: 159.
Facility Operating License No. NPF-42: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 23, 2004 (68
FR 68188).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 31, 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 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
[[Page 7776]]
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 email 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 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:
[[Page 7777]]
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 email 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).
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit 1, Berrien County, Michigan
Date of amendment request: January 15, 2005.
Description of amendment request: The amendment revises the
Operating License to add a license condition to allow a one-time
extension of the allowed outage time for the west centrifugal charging
pump.
Date of issuance: January 16, 2005.
Effective date: January 16, 2005.
Amendment No.: 285.
Facility Operating License No. DPR-58: Amendment revises the
Operating License.
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 16, 2005.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: M. Kotzalas, Acting.
Dated at Rockville, Maryland, this 7th day of February, 2005.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 05-2788 Filed 2-14-05; 8:45 am]
BILLING CODE 7590-01-P