[Federal Register Volume 69, Number 168 (Tuesday, August 31, 2004)]
[Notices]
[Pages 53098-53120]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-19586]
[[Page 53098]]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from, August 6 through August 19, 2004. The last
biweekly notice was published on August 19, 2004 (69 FR 51487).
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 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
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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 Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by email to
[email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: June 22, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification 3.1.8, ``Scram Discharge Volume (SDV)
Vent and Drain Valves,'' to allow a vent or drain line with one
inoperable valve to be isolated instead of requiring the valve to be
restored to Operable status within 7 days.
The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice
of opportunity for comment in the Federal Register on February 24, 2003
(68 FR 8637), on possible amendments to revise the action for one or
more SDV vent or drain lines with an inoperable valve, including a
model safety evaluation and model no significant hazards consideration
(NSHC) determination, using the consolidated line-item improvement
process. The NRC staff subsequently issued a notice of availability of
the models for referencing in license amendment applications in the
Federal Register on April 15, 2003 (68 FR 18294). The licensee affirmed
the applicability of the model NSHC determination in its application
dated June 22, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
A change is proposed to allow the affected SDV vent and drain
line to be isolated when there are one or more SDV vent or drain
lines with one valve inoperable instead of requiring the valve to be
restored to operable status within 7 days. With one SDV vent or
drain valve inoperable in one or more lines, the isolation function
would be maintained since the redundant valve in the affected line
would perform its safety function of isolating the SDV. Following
the completion of the required action, the isolation function is
fulfilled since the associated line is isolated. The ability to vent
and drain the SDV is maintained and controlled through
administrative controls. This requirement assures the reactor
protection system is not adversely affected by the inoperable
valves. With the safety functions of the valves being maintained,
the probability or consequences of an accident previously evaluated
are not significantly increased.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Accident Previously
Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. Thus,
this change does not create the possibility of a new or different
kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change ensures that the safety functions of the SDV
vent and drain valves are fulfilled. The isolation function is
maintained by redundant valves and by the required action to isolate
the affected line. The ability to vent and drain the SDV is
maintained through administrative controls. In addition, the reactor
protection system will prevent filling of the SDV to the point that
it has insufficient volume to accept a full scram. Maintaining the
safety functions related to isolation of the SDV and insertion of
control rods ensures that the proposed change does not involve a
significant reduction in the margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60666.
NRC Section Chief: Anthony J. Mendiola.
[[Page 53100]]
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: April 23, 2004.
Description of amendment request: The proposed amendment would
delete Technical Specification (TS) Section 6.16, ``Post-Accident
Sampling Programs NUREG 0737 (II.B.3, II-F.1.2),'' and the related
requirements to maintain a Post-Accident Sampling System (PASS).
Licensees were generally required to implement PASS upgrades as
described in NUREG-0737, ``Clarification of TMI [Three Mile Island]
Action Plan Requirements,'' and Regulatory Guide 1.97, Revision 3,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Access
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the NRC's lessons
learned from the accident that occurred at TMI Unit 2. Requirements
related to PASS were imposed by Order for many facilities and were
added to or included in the TSs for nuclear power reactors currently
licensed to operate. Lessons learned and improvements implemented over
the last 20 years have shown that the information obtained from PASS
can be readily obtained through other means or is of little use in the
assessment and mitigation of accident conditions.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on March 3, 2003 (68 FR 10052) on possible amendments
to eliminate PASS, including a model safety evaluation and model no
significant hazards consideration (NSHC) determination, using the
consolidated line item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in a
license amendment application in the Federal Register on May 13, 2003
(68 FR 25664). The licensee affirmed the applicability of the following
NSHC determination in its application dated April 23, 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 PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radioisotopes
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Thomas S. O'Neill, Associate General
Counsel, AmerGen Energy Company, LLC, 4300 Winfield Road, Warrenville,
IL 60555.
NRC Section Chief: Richard J. Laufer.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: July 26, 2004.
Description of amendments request: The proposed amendments would
delete requirements from the Technical Specifications (TS) to maintain
hydrogen recombiners and hydrogen and oxygen monitors. Licensees were
generally required to implement upgrades as described in NUREG-0737,
``Clarification of TMI [Three Mile Island] Action Plan Requirements,''
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions
During and Following an Accident.'' Implementation of these upgrades
was an outcome of the lessons learned from the accident that occurred
at TMI, Unit 2. Requirements related to combustible gas control were
imposed by Order for
[[Page 53101]]
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,
``Combustible gas control for nuclear power reactors,'' eliminated the
requirements for hydrogen recombiners and relaxed safety
classifications and licensee commitments to certain design and
qualification criteria for hydrogen and oxygen monitors.
The 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 26, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen and oxygen monitors are no longer required to mitigate
design-basis accidents and, therefore, the hydrogen monitors do not
meet the definition of a safety-related component as defined in 10
CFR 50.2. RG 1.97, Category 1, is intended for key variables that
most directly indicate the accomplishment of a safety function for
design-basis accident events. The hydrogen and oxygen monitors no
longer meet the definition of Category 1 in RG 1.97. As part of the
rulemaking to revise 10 CFR 50.44, the Commission found that
Category 3, as defined in RG 1.97, is an appropriate categorization
for the hydrogen monitors because the monitors are required to
diagnose the course of beyond design-basis accidents. Also, as part
of the rulemaking to revise 10 CFR 50.44, the Commission found that
Category 2, as defined in RG 1.97, is an appropriate categorization
for the oxygen monitors, because the monitors are required to verify
the status of the inert containment.
The regulatory requirements for the hydrogen and oxygen monitors
can be relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3,
classification of the oxygen monitors as Category 2 and removal of
the hydrogen and oxygen monitors from TS will not prevent an
accident management strategy through the use of the SAMGs [severe
accident management guidelines], the emergency plan (EP), the
emergency operating procedures (EOPs), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen and oxygen monitor
requirements, including removal of these requirements from TS, does
not involve a significant increase in the probability or the
consequences of any accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen and oxygen monitor requirements,
including removal of these requirements from TS, will not result in
any failure mode not previously analyzed. The hydrogen recombiner
and hydrogen and oxygen monitor equipment was intended to mitigate a
design-basis hydrogen release. The hydrogen recombiner and hydrogen
and oxygen monitor equipment are not considered accident precursors,
nor does their existence or elimination have any adverse impact on
the pre-accident state of the reactor core or post accident
confinement of radionuclides within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen and oxygen monitor requirements,
including removal of these requirements from TS, in light of
existing plant equipment, instrumentation, procedures, and programs
that provide effective mitigation of and recovery from reactor
accidents, results in a neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI, Unit 2, accident
can be adequately met without reliance on safety-related hydrogen
monitors. Category 2 oxygen monitors are adequate to verify the
status of an inerted containment.
Therefore, this change does not involve a significant reduction
in the margin of safety. The intent of the requirements established
as a result of the TMI, Unit 2, accident can be adequately met
without reliance on safety-related oxygen monitors. Removal of
hydrogen and oxygen monitoring from TS will not result in a
significant reduction in their functionality, reliability, and
availability.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief (Acting): Michael L. Marshall.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: June 21, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification Section 5.5.14, ``Technical
Specifications (TS) Bases Control Program,'' to replace the previous 10
CFR 50.59 term ``unreviewed safety question'' with current terminology.
The proposed amendment would also revise TS Section 5.7.1, ``High
Radiation Area,'' to add wording that was inadvertently deleted with
the issuance of the Improved Standard Technical Specifications in
Amendment No. 176.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
[[Page 53102]]
The proposed changes do not modify the facility or the
procedures for operation of the facility. One change updates the
terminology used in 10 CFR 50.59 evaluations. The change does not
alter the requirement of the TS Bases Control Program. The
requirement for NRC review and approval of a TS Bases change is
still determined through the use of the 10 CFR 50.59 review process.
The second change corrects a typographical error that occurred under
Amendment No. 176. The wording as proposed in this correction
restores the requirement to the phraseology approved in Amendment
No. 152 and is consistent with existing plant procedures.
Since there are no changes to the facility or facility
procedures, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed changes do not modify the facility or the
procedures for operation of the facility. One change updates the
terminology used in 10 CFR 50.59 evaluations. The change does not
alter the requirement of the TS Bases Control Program. The
requirement for NRC review and approval of a TS Bases change is
still determined through the use of the 10 CFR 50.59 review process.
The second change corrects a typographical error that occurred under
Amendment No. 176. The wording as proposed in this correction
restores the requirement to the phraseology approved in Amendment
No. 152 and is consistent with existing plant procedures.
Since there are no changes to the facility or facility
procedures, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The proposed changes continue to provide the controls necessary
to ensure changes to the TS Bases are made in conformance with 10
CFR 50.59. The proposed changes continue to provide the controls
necessary to ensure adequate control of High Radiation Areas. The
proposed changes will not result in any changes to the facility or
facility operating procedures. Therefore, the changes do not result
in a significant reduction in the margin of safety.
Based on the above discussion, Carolina Power & Light has
determined that the requested change does not involve a significant
hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Michael L. Marshall, Acting.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: June 9, 2004.
Description of amendment request: The proposed change revises
Technical Specifications (TS) Limiting Condition for Operation (LCO)
3.4.11, ``RCS Pressure and Temperature (P/T) Limits,'' to replace the
P/T curves for inservice leak and hydrostatic testing, non-nuclear
heating and cooldown, and nuclear heating and cooldown currently
illustrated in TS Figures 3.4.11-1, 3.4.11-2, and 3.4.11-3,
respectively.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes deal exclusively with the Reactor Coolant
System (RCS) Pressure and Temperature (P/T) curves, which define the
limitations for operation and testing. Because of the design
conservatisms used to calculate the RCS P/T limits, reactor vessel
failure has a low probability of occurrence and is not considered as
a design basis accident in the safety analyses of the plant. The
proposed changes adjust the reference temperature for the limiting
material to account for irradiation effects and provide a comparable
level of protection as previously evaluated and approved. The
adjusted reference temperature calculations were performed in
accordance with the requirements of 10 CFR [Part] 50 Appendix G
using the guidance contained in RG [Regulatory Guide] 1.99, Revision
2, ``Radiation Embrittlement of Reactor Vessel Materials,'' to
provide operating limits for up to 33.1 EFPY [effective full power
years]. The proposed license amendment does not involve a change to
operation of equipment required to mitigate any accident analyzed in
Columbia's UFSAR [Updated Final Safety Analysis Report]. Therefore,
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The revised P/T curves are based on a later edition and addenda
of the ASME Code that incorporates current industry standards for
the curves. The revised curves are also based on an RPV [reactor
pressure vessel] fluence that has been recalculated in accordance
with the methodology of RG 1.190. The proposed changes do not
involve a modification to plant equipment. There is no effect on the
function of any plant system, and no new system interactions are
introduced by this change. No new failure modes are introduced.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed curves conform to the guidance contained in RG
1.190, ``Calculational and Dosimetry Methods for Determining
Pressure Vessel Neutron Fluence,'' and RG 1.99, Revision 2,
``Radiation Embrittlement of Reactor Vessel Materials,'' and
maintain the safety margins specified in 10 CFR [Part] 50 Appendix
G. 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: Thomas C. Poindexter, Esq., Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: August 5, 2004.
Description of amendment request: The proposed change will revise
Technical Specification (TS) 5.5.12, ``Primary Containment Leakage Rate
Testing Program,'' to allow a one-time deferral of the Type A
containment integrated leak rate test (ILRT). The current 10-year
interval between Type A tests would be extended to 15 years from the
previous time a Type A test was performed. The last Type A test was
performed on July 20, 1994.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed one-time extension to the Type A testing interval
from once-per-10 years to once-per-15 years will not increase the
probability of an accident previously evaluated. The performance of
Type A tests is not an accident initiator. The primary containment
Type A testing interval extension does not involve a plant
[[Page 53103]]
modification and will not cause equipment failure or accident
initiation.
The proposed extension to the Type A testing interval does not
involve a significant increase in the consequences of an accident.
The NUREG 1493 generic study of the effects of extending containment
leakage testing concluded that Type B and C testing can identify the
vast majority (greater than 95 percent) of potential leakage paths
and that reducing the Type A test interval to once-per-20 years
leads to an ``imperceptible increase in risk.'' Other testing and
inspection programs, in addition to the Type A test, provide a high
degree of assurance that the primary containment integrity will be
maintained. Inspections required by the Maintenance Rule and ASME
Code [are] periodically performed in order to identify indications
of containment degradation that could affect containment leak
tightness.
Experience at Columbia demonstrates that excessive containment
leakage paths are detectable by Type B and C local leak rate tests.
Type B and C testing will identify containment openings, such as a
valve, that would otherwise be detected by the Type A test. These
factors show that a one-time Type A test interval extension from
once-per-10 years to once-per-15 years will not involve a
significant increase in the consequences of an accident.
Previous Type A test results at Columbia show leakage has not
exceeded acceptance criteria in the past, indicating a leak-tight
containment and demonstrating the structural capability of the
primary containment. The testing results have established that
Columbia has had acceptable containment leakage rates with
considerable margin.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The Columbia primary containment is designed to contain energy
and fission products during and after a design basis accident. The
proposed extension of the Type A testing interval will not create
the possibility of a new or different type of accident from any
previously evaluated. There are no changes being made to the
physical plant or in operation of the plant that could introduce a
new failure mode with the potential to create an accident or affect
mitigation of an accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed extension of the Type A testing interval will not
significantly reduce the margin of safety. The NUREG 1493 generic
study of the effects of extending containment leakage testing found
that a 20-year interval in Type A leakage testing leads to an
``imperceptible increase in risk.'' NUREG 1493 found that
generically, the design containment leakage rate contributes less
than 0.1 percent to the overall accident risk and that the increase
in the Type A testing interval would have a minimal effect on risk
because the vast majority (greater than 95 percent) of all potential
leakage paths are detected by Type B and C leakage testing.
A Columbia plant specific probabilistic risk assessment on the
change in the Type A test interval from once-per-10 years to once-
per-15 years determined:
The risk impact due to a change in Large Early Release
Frequency (LERF) is an increase of 2E-8/year that is characterized
by Regulatory Guide 1.174 [``An Approach for Using Probabilistic
Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes
to the Licensing Basis''] as ``very small.''
The total integrated plant risk increase measured by
person-rem/year is negligible.
The change in conditional containment failure
probability is an increase of 0.1 percent, which is considered to
represent a very small impact on risk.
Deferral of Type A testing for Columbia does not increase the
level of risk to the public due to loss of capability to detect and
measure containment leakage or loss of containment structural
integrity. Other containment testing methods and inspections will
assure all limiting conditions for operation will continue to be
met. The margin of safety inherent in existing accident analyses
will be maintained.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Thomas C. Poindexter, Esq., Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: June 22, 2004.
Description of amendment request: The proposed amendment would
delete requirements from the Technical Specifications (TSs) to maintain
hydrogen and oxygen monitors. A notice of availability for this
technical specification improvement using the consolidated line item
improvement process (CLIIP) was published in the Federal Register (FR)
on September 25, 2003 (68 FR 55416). Licensees were generally required
to implement upgrades as described in NUREG-0737, ``Clarification of
TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory
Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power
Plants to Assess Plant and Environs Conditions During and Following an
Accident.'' 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 (not installed at FitzPatrick and therefore not addressed
by this proposed amendment) 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 FR on September 25, 2003 (68
FR 55416). The licensee affirmed the applicability of the model NSHC
determination in its application dated June 22, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen and oxygen monitors are no longer required to mitigate
design-basis accidents and, therefore, the hydrogen monitors do not
meet the definition of a safety-related component as defined in 10
CFR 50.2. RG [Regulatory Guide] 1.97 Category 1, is intended for key
[[Page 53104]]
variables that most directly indicate the accomplishment of a safety
function for design-basis accident events. The hydrogen and oxygen
monitors no longer meet the definition of Category 1 in RG 1.97. As
part of the rulemaking to revise 10 CFR 50.44 the Commission found
that Category 3, as defined in RG 1.97, is an appropriate
categorization for the hydrogen monitors because the monitors are
required to diagnose the course of beyond design-basis accidents.
Also, as part of the rulemaking to revise 10 CFR 50.44, the
Commission found that Category 2, as defined in RG 1.97, is an
appropriate categorization for the oxygen monitors, because the
monitors are required to verify the status of the inert containment.
The regulatory requirements for the hydrogen and oxygen monitors
can be relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3,
[classification of the oxygen monitors as Category 2,] and removal
of the hydrogen and oxygen monitors from TS will not prevent an
accident management strategy through the use of the severe accident
management guidelines (SAMGs), the emergency plan (EP), the
emergency operating procedures (EOPs), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the relaxation of the hydrogen and oxygen monitor
requirements, including removal of these requirements from TS, does
not involve a significant increase in the probability or the
consequences of any accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The relaxation of the hydrogen and oxygen monitor requirements,
including removal of these requirements from TS, will not result in
any failure mode not previously analyzed. The hydrogen and oxygen
monitor equipment was intended to mitigate a design-basis hydrogen
release. The hydrogen and oxygen monitor equipment are not
considered accident precursors, nor does their existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within
the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The relaxation of the hydrogen and oxygen monitor requirements,
including removal of these requirements from TS, in light of
existing plant equipment, instrumentation, procedures, and programs
that provide effective mitigation of and recovery from reactor
accidents, results in a neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI, Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Category 2 oxygen monitors are adequate to verify the status of
an inerted containment.
Therefore, this change does not involve a significant reduction
in the margin of safety. The intent of the requirements established
as a result of the TMI, Unit 2 accident can be adequately met
without reliance on safety-related oxygen monitors. Removal of
hydrogen and oxygen monitoring from TS will not result in a
significant reduction in their functionality, reliability, and
availability.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: June 2, 2004.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to fully adopt the alternate
source term (AST) methodology for design-basis accident dose
consequence evaluations in accordance with 10 CFR 50.67. Specifically,
the amendment would revise the TS Definition regarding dose equivalent
iodine and TS Section 5.5.10, ``Ventilation Filter Testing Program
(VFTP).'' The AST methodology for the fuel-handling accident was
previously approved in Amendment No. 215, dated March 17, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change involves the reanalysis of design basis
radiological accidents in Containment and the Fuel Storage Building.
The new analyses, based on the Alternate Source Term (AST), in
accordance with 10 CFR 50.67, will replace the existing analyses
that are based on the methodologies of [Atomic Energy Commission
Report, ``Calculation of Distance Factors for Power and Test Reactor
Sites,'' 1962] TID-14844. As a result of the new analyses, changes
to the Technical Specifications are proposed which take credit for
the new analysis results.
The proposed changes to the Technical Specifications modify
requirements regarding filter testing for a variety of systems
(i.e., Containment Purge, Fuel Storage Building Emergency
Ventilation). The analyses do not credit charcoal or HEPA [high-
efficiency particulate air] filtration for dose mitigation. The
proposed changes reflect the plant configuration that will support
implementation of the AST analyses.
The AST analysis follows the guidance of the NRC Regulatory
Guide 1.183 and uses the acceptance criteria of the NRC Standard
Review Plan (NUREG-0800) for offsite doses and General Design
Criteria for Control Room personnel. The accident analyses
conservatively assume that the Containment Building and the Fuel
Storage Building, including ventilation filtration systems for those
buildings, do not diminish or delay the assumed fission product
release.
The proposed changes also revise the definition of Dose
Equivalent Iodine (DEI) to be consistent with the assumptions of the
analyses. The limits for DEI do not change as a result of the
implementation of the AST analyses.
The change from the original source term to the new proposed AST
is a change in analysis method and assumptions and has no effect on
accident initiators or causal factors that contribute to the
probability of occurrence of previously analyzed accidents. Use of
AST to analyze the dose effect of design basis accidents shows that
regulatory acceptance criteria for the new methodology continue to
be met. Changing the analysis methodology does not change the
sequence or progression of the accident scenario.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The changes proposed in this license amendment request involve
the use of a new analysis methodology and related regulatory
acceptance criteria. In addition, certain changes to plant
ventilation systems can be made based on the analysis results, using
the new methodology. Use of a new analysis
[[Page 53105]]
method does not impact the design or operation of plant systems or
components and new accident scenarios would therefore not be
created. The proposed changes to air ventilation and filtration
systems do not adversely affect plant equipment used to protect
plant safety limits or the way in which that plant equipment is
operated or maintained. As a result, no new failure modes are being
introduced that could lead to different accidents.
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 existing dose analysis methodology and assumptions
demonstrate that the dose consequences for all design basis
accidents are within regulatory limits for whole body and thyroid
doses as established in 10 CFR 100 (except for the Fuel Handling
Analysis, which is already based on the AST methodology). The
alternate dose analysis methodology and assumptions also demonstrate
that the dose consequences of these accidents are within the
regulatory requirements established for the new methodology.
The limits applicable to the alternate analysis are established
in 10 CFR 50.67 in conjunction with the Total Effective Dose
Equivalent (TEDE) acceptance directed in Regulatory Guide 1.183. The
acceptance criteria for both dose analysis methods have been
developed for the purpose of evaluating design basis accidents to
demonstrate adequate protection of public health and safety. An
acceptable margin of safety is inherent in both types of acceptance
criteria.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: June 3, 2004.
Description of amendment request: The proposed amendment would
increase the maximum authorized reactor core power level from 3067.4
megawatt thermal (MWt) to 3216 MWt. This represents a nominal increase
of 4.85% rated thermal power. The amendment would also revise the
Technical Specifications (TSs) to relocate certain cycle-specific
parameters to the Core Operating Limits Report (COLR) by adopting TS
Task Force Traveler TSTF-339, ``Relocate Technical Specification
Parameters to the COLR.'' In addition, the amendment would revise
several allowable values in TS Table 3.3.1-1, ``Reactor Protection
System (RPS) Instrumentation,'' and Table 3.3.2-1, ``Engineered Safety
Feature Actuation System (ESFAS) Instrumentation.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The evaluations and analyses associated with this proposed
change to core power level have demonstrated that all applicable
acceptance criteria for plant systems, components, and analyses
(including the Final Safety Analysis Report Chapter 14 safety
analyses) will continue to be met for the proposed increase in
licensed core thermal power for Indian Point 3 (IP3). The subject
increase in core thermal power will not result in conditions that
could adversely affect the integrity (material, design, and
construction standards) or the operational performance of any
potentially affected system, component or analysis. Therefore, the
probability of an accident previously evaluated is not affected by
this change. The subject increase in core thermal power will not
adversely affect the ability of any safety-related system to meet
its intended safety function. Further, the radiological dose
evaluations in support of this power uprate effort show all
acceptance criteria are met.
The relocation of cycle-specific core operating limits from the
Technical Specifications to the Core Operating Limits Report (COLR),
in accordance with TSTF-339, has no influence or impact on the
probability or consequences of a Design Basis Accident. Adherence to
the COLR and accepted methodologies for establishing COLR parameters
continues to be controlled by the plant Technical Specifications.
Relocation of cycle-specific values to the COLR while maintaining
the limiting requirements in the Technical Specifications reduces
administrative burden associated with processing license amendments
for routine core reload designs.
RPS and ESF [engineered safety feature] allowable values
established in plant technical specifications represent acceptance
criteria used by plant personnel in assessing the operability of
instrumentation channels.
Allowable values are not accident initiators and have no role in
the probability of occurrence of an accident. Safety analyses for
design basis accidents use certain assumptions (Safety Analysis
Limits) regarding the actuation of RPS and ESF protective functions.
The proposed allowable values are developed using a methodology that
assures the accident analysis assumptions are valid and the
consequences of previously analyzed accidents continue to meet
established limits.
Therefore, the proposed changes described in this license
amendment request do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The analyses and evaluations performed for the proposed increase
in power show that all applicable acceptance criteria for plant
systems, components, and analyses (including FSAR [Final Safety
Analysis Report] Chapter 14 safety analyses) will continue to be met
for the proposed power increase in IP3 licensed core thermal power.
The subject increase in core thermal power will not result in
conditions that could adversely affect the integrity (material,
design, and construction standards) or operational performance of
any potentially affected system, component, or analyses. The subject
increase in core thermal power will not adversely affect the ability
of any safety-related system to meet its safety function.
Furthermore, the conditions and changes associated with the subject
increase in core thermal power will neither cause initiation of any
accident, nor create any new credible limiting single failure. The
power uprate does not result in changing the status of events
previously deemed to be non-credible being made credible.
Additionally, no new operating modes are proposed for the plant as a
result of this requested change.
The relocation of cycle-specific core operating limits from the
Technical Specifications to the Core Operating Limits Report (COLR),
in accordance with TSTF-339, does not involve any changes to plant
equipment or the way is which the plant is operated. There are no
new accident initiators or causal mechanisms being introduced by
this proposed change. Relocation of cycle-specific values to the
COLR while maintaining the limiting requirements in the Technical
Specifications reduces administrative burden associated with
processing license amendments for routine core reload designs.
RPS and ESF allowable values established in plant technical
specifications represent acceptance criteria used by plant personnel
in assessing the operability of instrumentation channels. Revising
allowable values does not involve installation of new equipment,
modification to existing equipment, or a change in plant operation
that could create a new or different accident scenario.
Therefore, the proposed changes described in this license
amendment request will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
[[Page 53106]]
Response: No.
The analyses and evaluations associated with the proposed
increase in power show that all applicable acceptance criteria for
plant systems, components, and analyses (including FSAR Chapter 14
safety analyses) will continue to be met for this proposed increase
in IP3 licensed core thermal power. The subject increase in core
thermal power will not result in conditions that could adversely
affect the integrity (material, design, and construction standards)
or operational performance of any potentially affected system,
component, or analysis. The subject power uprate will not adversely
affect the ability of any safety-related system to meet its intended
safety function.
Adoption of TSTF-339 allows relocation of cycle-specific
parameters to the COLR, while maintaining limiting requirements in
the Technical Specifications. Approved methodologies for calculating
cycle-specific parameters are maintained in the Technical
Specifications, and changes to the COLR are subject to the
requirements and controls of 10 CFR 50.59. This assures that
required margins to safety limits are maintained.
The proposed new allowable values are developed using
established methodologies and incorporate additional conservatism
that assures the validity of analysis limits assumed in the
evaluation of hypothetical accidents.
Therefore, the proposed changes described in this license
amendment request 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: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: July 8, 2004.
Description of amendment request: Delete Technical Specification
Surveillance Requirement 4.5.2.d.1, Emergency Core Cooling System
Subsystems -Tave >= 300 [deg]F, associated with the
requirement to maintain an operable Automatic Closure Interlock (ACI)
for the Shutdown Cooling (SDC) suction isolation valves.
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 removal of the ACI function is consistent with the
guidelines previously endorsed by the NRC in Generic Letter 88-17.
Removal of this function results in a calculated decrease in
intersystem Loss of Coolant Accident (ISLOCA) frequency.
Additionally, the removal of the ACI function will result in a
decrease in SDC system unavailability and a corresponding decrease
in risk associated with loss of SDC events. As a result, the
proposed change will result in a net decrease in risk and a net
improvement in plant safety.
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 presence or omission of an ACI function is not considered an
accident initiator nor is this function credited in any safety
analyses for the prevention or mitigation of any accident. Alarms,
design features, and strict administrative/procedural controls
support correct and timely operator action to ensure the SDC system
will not be exposed to high Reactor Coolant System (RCS) pressure.
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 ACI function is not credited in a margin of safety analysis
for any accident previously evaluated. Removal of the ACI function
will result in an overall net increase in nuclear safety.
Appropriate alarm, design features, and administrative controls will
continue to ensure proper isolation and isolation maintenance of the
SDC system during plant operations with elevated RCS pressures.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: July 8, 2004. This supersedes the May
12, 2004, application in its entirety (69 FR 34699).
Description of amendment request: The proposed amendment would
change the reactor core analytical methods used to determine the core
operating limits, reflect the changes allowed by Technical
Specification (TS) Task Force (TSTF) Traveler No. 363, ``Revised
Topical Report References in ITS [Improved Standard Technical
Specifications] 5.6.5, COLR [Core Operating Limits Report],'' and
delete the Index from the TSs. This request completely supersedes the
previous request of May 12, 2004.
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.
TS 6.9.5.1, Core Operating Limits Report (COLR)
The proposed amendment, in part, identifies a change in the
nuclear physics codes used to confirm the values of selected cycle-
specific reactor physics parameter limits and includes minor
editorial changes which do not alter the intent of stated
requirements. The proposed change also allows the use of methods
required for the implementation of ZIRLO clad fuel rods. Inasmuch as
the proposed change includes codes that have been previously
approved by the NRC for CE [Combustion Engineering] cores, the
amendment is administrative in nature and has no impact on any plant
configuration or system performance relied upon to mitigate the
consequences of an accident. Parameter limits specified in the COLR
for this amendment are not changed from the values presently
required by TSs. Future changes to the calculated values of such
limits may only be made using NRC approved methodologies, must be
consistent with all applicable safety analysis limits, and are
controlled by the 10 CFR 50.59 process. Assumptions used for
accident initiators and/or safety analysis acceptance criteria are
not altered by this change.
The proposed change will add an NRC approved topical report,
WCAP-16072-P-A, to the list of referenced topical reports. The
topical report has been previously approved by the NRC for use in
Combustion Engineering core designs and as such, the proposed change
is administrative in nature and has no impact on any plant
configurations or on system performance that is relied upon to
mitigate the consequences of an accident. In addition, prior to the
use
[[Page 53107]]
of the ZrB2 burnable absorber coating, fuel design will
be analyzed with applicable NRC staff approved codes and methods.
The proposed change also implements NRC approved TSTF Traveler
No. 363. This is an administrative change that will allow specific
details, such as the revision number, revision date, and supplement
number of topical reports that are referenced in the TSs, to be
deleted and relocated in the cycle specific COLR. This proposed
change does not result in any changes to the assumptions used to
evaluated [evaluate] accident initiators and/or safety analysis
acceptance criteria.
Index
The proposed deletion of the Index is purely administrative and
does not impact the accident analysis.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
TS 6.9.5.1, Core Operating Limits Report (COLR)
The proposed change, in part, identifies a change in the nuclear
physics codes used to confirm the values of selected cycle-specific
reactor physics parameter limits. The proposed change also allows
the use of methods required for the implementation of ZIRLO clad
fuel rods. Neither of these changes results in a change to the
physical plant or to the modes of operation defined in the facility
license.
The proposed change adds a reference to the topical report that
allows the use of ZrB2 as a burnable absorber coating on
the fuel pellet. The topical report has been previously approved by
the NRC for use in Combustion Engineering core designs and as such,
the proposed change is administrative in nature and has no impact on
any plant configurations or on system performance that is relied
upon to mitigate the consequences of an accident. In addition, prior
to the use of the ZrB2 burnable absorber coating, fuel
design will be analyzed with applicable NRC staff approved codes and
methods. This change is administrative in nature and does not create
a new or different type of accident than previously evaluated
because the design requirements for the facility remain the same.
The proposed change also implements TSTF Traveler No. 363. The
proposed change does not result in changes to the physical plant or
to the modes of operation defined in the facility license nor does
it involve the addition of new equipment or the modification of
existing equipment.
Index
The proposed deletion of the Index is purely administrative has
no affect on existing equipment.
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.
TS 6.9.5.1, Core Operating Limits Report (COLR)
The proposed changes to change the nuclear physics code package
and to add a topical report to support the use of ZIRLO do not amend
the cycle specific parameter limits located in the COLR from the
values presently required by the TS. The individual specifications
continue to require operation of the plant within the bounds of the
limits specified in COLR. Benchmarking has shown that uncertainties
for the Westinghouse Physics code system yields are essentially the
same or less than those obtained for the current ROCS and DIT
[computer code] methodology. Future changes to the values of these
limits by the licensee may only be developed using NRC approved
methodologies, must remain consistent with all applicable plant
safety analysis limits addressed in the Safety Analysis Report, and
are further controlled by the 10 CFR 50.59 process. The relocation
of the supplement numbers, revision numbers, and approval dates of
the analytical methods listed in the COLR does not affect the margin
of safety. The analysis will continue to be performed using NRC
approved methodology. Safety analysis acceptance criteria are not
being altered by this amendment.
The proposed change will add WCAP-16072-P-A to the list of
referenced topical reports. The topical report has been previously
approved by the NRC for use in Combustion Engineering core designs
and as such, the proposed change is administrative in nature and has
no impact on any plant configurations or on system performance that
is relied upon to mitigate the consequences of an accident. In
addition, prior to the use of the ZrB2 burnable absorber
coating, fuel design will be analyzed with applicable NRC staff
approved codes and methods.
Index
The proposed deletion of the Index, which is an administrative
document, does not impact any TS values or safety limits.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: June 10, 2004, as supplemented by letter
dated July 21, 2004.
Description of amendment request: The proposed amendments would
revise the Quad Cities Nuclear Power Station (QCNPS) technical
specifications (TS) to change the allowable value (AV) and add
surveillance requirements (SRs) for the main steam line (MSL) flow-high
initiation of Group 1 primary containment isolation and control room
emergency ventilation system isolation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated?
For QCNPS, Units 1 and 2, the proposed amendment will implement
a design change that upgrades the existing MSL Flow-High
instrumentation from pressure switches to analog trip unit devices.
Analog trip units (ATUs) have proven to be a more reliable
technology than the currently installed equipment. Analog trip units
are used in various applications at QCNPS, including the Reactor
Protection System (RPS) low water level trip function. Because the
trip units are more reliable, the likelihood of spurious isolations
is reduced. Further, ATUs experience less instrument drift during
the operating cycle. The proposed change adds a 92-day trip unit
calibration requirement for the MSL-High isolation function. The NRC
has previously found that a 92-day calibration is appropriate for
individual ATUs.
Procedure revisions required by this modification are limited to
those associated with the calibration, maintenance, and operation of
the replacement transmitter and trip unit analog loops. All required
design functions of the MSL high flow loop are maintained. No
system, structure, or component will be used in a manner that is not
already bounded by the reference design, or is inconsistent with
analyses or descriptions in the QCNPS Updated Final Safety Analysis
Report (UFSAR). There is no adverse effect on the performance or
control of any design function described in the UFSAR.
TS requirements that govern operability or routine testing of
plant instruments are not assumed to be initiators of any analyzed
event because these instruments are intended to prevent, detect, or
mitigate accidents. Therefore, these changes will not involve an
increase in the probability of occurrence of an accident previously
evaluated. In addition, these changes will not increase the
[[Page 53108]]
consequences of an accident previously evaluated because the
proposed change does not adversely impact structures, systems, or
components. The planned instrument upgrade is a more reliable design
than existing equipment. The proposed changes establish requirements
that ensure components are operable when necessary for the
prevention or mitigation of accidents or transients. Furthermore,
there will be no change in the types or significant increase in the
amounts of any effluents released offsite. For these reasons, the
proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes support a planned instrumentation upgrade
by incorporating SRs required to ensure operability. The change does
not adversely impact the manner in which the instrument will operate
under normal and abnormal operating conditions. Therefore, these
changes provide an equivalent level of safety and will not create
the possibility of a new or different kind of accident from any
accident previously evaluated. The changes in methods governing
normal plant operation are consistent with the current safety
analysis assumptions.
All required design functions are maintained, and the new
setpoint is analyzed in accordance [with] an NRC-approved
methodology for determination of setpoints and TS AVs in accordance
with the QCNPS UFSAR, Section 7.3.2.4, ``Design Evaluation.''
Therefore, replacing the existing MSL high flow DPISs with analog
trip instrumentation does not alter any UFSAR described evaluation
methodologies, or introduce any new methodologies. These changes
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed changes support a planned instrumentation upgrade
from differential pressure switches to ATUs. The proposed changes do
not adversely affect the probability of failure or availability of
the affected instrumentation. The addition of a 92-day trip unit
calibration for MSL Flow-High is a conservative change that aligns
the SRs for a planned instrumentation upgrade with that of similar
instrumentation. The NRC has previously found that a 92-day
calibration is appropriate for individual ATUs. The setpoint was
determined using an NRC-approved methodology. The proposed changes
do not affect the analytical limit assumed in the safety analyses
for the actuation of the instrumentation. Therefore, it is concluded
that the proposed changes will not result in a reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of amendment request: March 22, 2004 as supplemented July 23,
2004.
Description of amendment request: The proposed change allows entry
into a mode or other specified condition in the applicability of a
technical specification (TS), while in a condition statement and the
associated required actions of the TS, provided the licensee performs a
risk assessment and manages risk consistent with the program in place
for complying with the requirements of Title 10 of the Code of Federal
Regulations (10 CFR), part 50, Section 50.65(a)(4). Limiting Condition
for Operation (LCO) 3.0.4 exceptions in individual TSs would be
eliminated, several notes or specific exceptions are revised to reflect
the related changes to LCO 3.0.4, and Surveillance Requirement (SR)
4.0.4 is revised to reflect the LCO 3.0.4 allowance.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a
notice of opportunity for comment in the Federal Register on August 2,
2002 (67 FR 50475), on possible amendments concerning TSTF-359,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on April 4, 2003 (68 FR 16579).
The licensee affirmed the applicability of the following NSHC
determination in its application dated March 22, 2004 and July 23,
2004, supplement.
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 proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed by proposed LCO 3.0.4, are no different than the
consequences of an accident while entering and relying on the
required actions while starting in a condition of applicability of
the TS. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. Thus, this change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. The TS
allow operation of the plant without the full complement of
equipment through the conditions for not meeting the TS LCO. The
risk associated with this allowance is managed by the imposition of
required actions that must be performed within the prescribed
completion times. The net effect of being in a TS condition on the
margin of safety is not considered significant. The proposed change
does not alter the required actions or completion times of the TS.
The proposed change allows TS conditions to be entered, and the
associated required actions and completion times to be used in new
circumstances. This use is predicated upon the licensee's
performance of a risk assessment and the management of plant risk.
The change also eliminates current allowances for utilizing required
actions and completion times in similar circumstances, without
assessing and managing risk. The net change to the margin of safety
is insignificant. Therefore, this change does not involve a
significant reduction in a margin of safety.
[[Page 53109]]
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412,
Beaver Valley Power Station, Unit No. 2 (BVPS-2), Beaver County,
Pennsylvania
Date of amendment request: July 23, 2004.
Description of amendment request: The proposed amendment would
revise the BVPS-2 Technical Specifications to eliminate periodic
response time testing requirements on selected sensors and selected
protection channel components and permit the option of measuring or
verifying the response times by means other than testing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This change to the Technical Specifications does not result in a
condition where the design, material, and construction standards
that were applicable prior to the change are altered. The same RTS
[reactor trip system] and ESFAS [engineered safety features
actuation system] instrumentation is being used; the time response
allocations/modeling assumptions in the Updated Final Safety
Analysis Report (UFSAR) Chapter 15 analyses are still the same; only
the method of verifying [the] time response is changed. The proposed
change will not modify any system interface and could not increase
the likelihood of an accident since these events are independent of
this change. The proposed activity will not change, degrade or
prevent actions or alter any assumptions previously made in
evaluating the radiological consequences of an accident described in
the UFSAR.
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.
This change does not alter the performance of the pressure and
differential pressure transmitters, process protection racks,
Nuclear Instrumentation, and logic systems used in the Reactor Trip
and Engineered Safety Features Actuation Systems. All sensors,
process protection racks, Nuclear Instrumentation, and logic systems
will still have response time verified by [a] test before placing
the equipment into operational service and after any maintenance
that could affect the response time. Changing the method of
periodically verifying instrument response times for certain
equipment (assuring equipment operability) from time response
testing to calibration and channel checks will not create any new
accident initiators or scenarios. Periodic surveillance of these
instruments will detect significant degradation in the equipment
response time characteristics.
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.
This change does not affect the total system response time
assumed in the safety analysis. The periodic system response time
verification method for selected sensors and differential pressure
sensors and for process protection racks, Nuclear Instrumentation,
and logic systems is modified to allow use of actual test data or
engineering data. The method of verification still provides
assurance that the total system response time is within that assumed
in the safety analysis.
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 O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: April 26, 2004.
Description of amendment request: This proposed license amendment
would revise the frequency of the Mode 5 Intermediate Range Monitoring
(IRM) Instrumentation CHANNEL FUNCTIONAL TEST contained in Technical
Specification (TS) 3.3.1.1 from 7 days to 31 days. The methodology used
to analyze the change in testing frequency is based upon guidance
contained in Generic Letter 91-04, ``Changes in Technical Specification
Surveillance Intervals to Accommodate a 24-month Fuel Cycle,'' and
Electric Power Institute (EPRI) Report TI-103335, ``Guidance for
Instrumentation Calibration Extension/Reduction Programs.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed Technical Specification (TS) change involves an
increase in the Mode 5 CHANNEL FUNCTIONAL TEST interval for Reactor
Protection System (RPS) Intermediate Range Monitor (IRM) from 7 days
to 31 days. The proposed TS change does not alter the design or
functional requirements of the RPS or IRM systems. Evaluation of the
proposed testing interval change demonstrated that the availability
of the IRMs to prevent or mitigate the consequences of a control rod
withdrawal event at low power levels are not significantly affected
because of other, more frequent testing that is performed, the
availability of redundant systems and equipment, and the high
reliability of the IRM equipment.
Furthermore, using the guidance of GL 91-04, a historical review
of surveillance test results and associated maintenance records did
not indicate evidence of any failure that would invalidate the above
conclusions.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed TS change involves an increase in the Mode 5 IRM
CHANNEL FUNCTIONAL TEST interval from 7 days to 31 days. Existing TS
testing requirements ensure the operability of the IRMs. The
proposed TS change does not introduce any failure mechanisms of a
different type than those previously evaluated, since no physical
changes to the plant are being made. No new or different equipment
is being installed, and no installed equipment is being operated in
a different manner. As a result, no new failure modes are
introduced. In addition, the manner in which surveillance tests are
performed remain unchanged.
Furthermore, using the guidance in GL 91-04, a historical review
of surveillance test results and associated maintenance records did
not indicate evidence of any failure that would invalidate the above
conclusions.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change will not involve a single reduction in
the margin of safety.
The proposed Technical Specifications (TS) change involves an
increase in the Mode 5 CHANNEL FUNCTIONAL TEST interval for Reactor
Protection System (RPS) Intermediate Range Monitor (IRM) from 7 days
to 31 days. The impact on system operability is minimal, based upon
performance of the more frequent Channel
[[Page 53110]]
Checks, continuous Control Room monitoring when the IRMs are in use,
and the overall IRM reliability. Evaluations show there is no
evidence of time-dependent failures that would impact the
availability of the IRMs.
Furthermore, using the guidance in GL 91-04, a historical review
of surveillance test results and associated maintenance records did
not indicate evidence of any failure that would invalidate the above
conclusions.
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: Anthony J. Mendiola.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: June 28, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification 3/4.9.4, ``Containment Building
Penetrations,'' to align the language of the Surveillance Requirement
with the Applicability Statement contained in the Limiting Condition
for Operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change aligns the language of the Surveillance
Requirement for Containment Building Penetrations with the language
of the Applicability Statement of Technical Specification 3.9.4.
The proposed amendment will not change the design function, or
method of performing or controlling design functions, of structures,
systems and components, nor will there be an effect on FPL Energy
Seabrook programs. As a result, the proposed amendment will not
change assumptions, or change, degrade or prevent actions described
or assumed in accidents evaluated and described in the Seabrook
Station UFSAR [updated final safety analysis report]. The proposed
change to the Surveillance Requirement wording does not adversely
affect performance of the Surveillance Requirement that verifies the
status of Containment Building Penetrations. Since the status of the
Containment Penetrations is not adversely affected by the proposed
change, the radiological consequences of an event are unchanged.
Therefore, the proposed amendment does not result in an increase in
the radiological consequences of any accident described in the
Seabrook Station UFSAR.
Therefore, it is concluded that these proposed changes do not
involve a significant increase in the probability or consequence of
an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change aligns the language of the Surveillance
Requirements for Containment Building Penetrations with the language
in the Applicability Statement of the Technical Specification.
The proposed amendment will not change the design function, or
method of performing or controlling design functions, of structures,
systems and components, nor will there be an effect on FPL Energy
Seabrook programs. As a result, there are no changes associated with
the proposed amendment that could potentially introduce new failure
modes or accident scenarios.
Therefore, it is concluded that these proposed changes do not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The proposed change aligns the language of the Surveillance
Requirement for Containment Building Penetrations with the language
of the Applicability Statement of Technical Specification 3.9.4. The
proposed amendment does not change the design function, or method of
performing or controlling design functions, of structures, systems
and components, nor will there be an effect on FPL Energy Seabrook
programs. The status of containment penetrations will continue to be
verified. The proposed change does not involve any changes to a
margin of safety.
Therefore, it is concluded that these proposed changes 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: M. S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Section Chief: James W. Clifford.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: August 17, 2004.
Description of amendment request: The licensee proposed to revise
Section 3.3.1, ``Oxygen Concentration [of the primary containment],''
of the Technical Specifications (TSs) to (1) add a new action allowing
24 hours to restore the oxygen concentration to within the limit of <4%
by volume if the limit is exceeded when the reactor is in the power
operating condition, and (2) incorporate the associated conforming
changes of editorial nature. The proposed 24-hour completion time for
restoring oxygen concentration is consistent with Improved Standard
Technical Specifications for Boiling Water Reactors (NUREG-1433,
Revision 3).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff's analysis is presented below:
The first standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
increase in the probability or consequences of an accident previously
evaluated. The effect of the proposed amendment is to provide the same
24-hour completion time to restore oxygen concentration to under the 4%
limit should the oxygen concentration rise due to other than a reactor
shutdown-startup evolution. The proposed amendment does not lead to,
nor is it the result of, a plant design change. These TS changes will
not lead to alteration of the physical design or operational procedures
associated with the containment system, or any other plant structure,
system, or component (SSC). All requirements needed to assure
operability of the containment system will remain unchanged.
Containment atmospheric oxygen concentration was not assumed to be a
precursor of accidents, nor was it assumed to be a component in
previously evaluated accident scenarios. Accordingly, the revised
specifications will lead to no increase in the consequences of an
accident previously evaluated, and no increase of the probability of an
accident previously evaluated.
The second standard requires that operation of the unit in
accordance with the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
evaluated. As stated above, the proposed amendment involves only the
time allowed to restore containment atmospheric oxygen concentration to
under 4 percent by volume, and associated editorial changes. These
[[Page 53111]]
changes do not alter the physical design, safety limits, or method of
operation associated with the operation of the plant. Accordingly, the
changes do not introduce any new or different kind of accident from
those previously evaluated.
The third standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
reduction in a margin of safety. Since the licensee did not propose to
exceed or alter a design basis or safety limit, did not propose to
operate any component in a less conservative manner, and did not
propose to use a less conservative analysis methodology, the proposed
amendment will not affect in any way the performance characteristics
and intended functions of any SSC. Therefore, the proposed amendment
does not involve a significant reduction in a margin of safety.
Based on the NRC staff's analysis, it appears that the three
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the proposed amendment 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.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of amendment request: July 6, 2004.
Description of amendment request: The proposed change involves the
extension from 1 hour to 24 hours for the completion time (CT) of
Technical Specification (TS) 3.3.a.2.B, which defines requirements for
accumulators. Accumulators are part of the emergency core cooling
system and consist of tanks partially filled with borated water and
pressurized with nitrogen gas. The contents of the tank are discharged
to the reactor coolant system (RCS) if, as during a loss-of-coolant
accident, the coolant pressure decreases to below the accumulator
pressure. TS 3.3.a.2.B specifies a CT to restore an accumulator to
operable status when it has been declared inoperable for a reason other
than the boron concentration of the water in the accumulator not being
within the required range. This change was proposed by the Westinghouse
Owners Group participants in the TS Task Force (TSTF) and is designated
TSTF-370, ``Increase Accumulator Completion Time from 1 Hour to 24
Hours.'' TSTF-370 is supported by NRC-approved Topical Report WCAP-
15049-A, ``Risk-Informed Evaluation of an Extension to Accumulator
Completion Times,'' submitted on May 18, 1999. The NRC staff issued a
notice of opportunity for comment in the Federal Register on July 15,
2002 (67 FR 46542), on possible amendments concerning TSTF-370,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line-item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on March 12, 2003 (68 FR 11880).
The licensee affirmed the applicability of the following NSHC
determination in its application dated July 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:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The basis for the accumulator limiting condition for operation
(LCO), as discussed in [Standard Technical Specifications] Bases
Section 3.5.1, is to ensure that a sufficient volume of borated
water will be immediately forced into the core through each of the
cold legs in the event the RCS pressure falls below the pressure of
the accumulators, thereby providing the initial cooling mechanism
during large RCS pipe ruptures. As described in Section 9.2 of the
WCAP-15049, ``Risk-Informed Evaluation of an Extension to
Accumulator Completion Times,'' evaluation, the proposed change will
allow plant operation in a configuration outside the design basis
for up to 24 hours, instead of 1 hour, before being required to
begin shutdown. The impact of the increase in the accumulator CT on
core damage frequency for all the cases evaluated in WCAP-15049 is
within the acceptance limit of 1.0E-06/yr for a total plant core
damage frequency (CDF) less than 1.0E-03/yr. The incremental
conditional core damage probabilities calculated in WCAP-15049 for
the accumulator CT increase meet the criterion of 5E-07 in
Regulatory Guides (RG) 1.174 and 1.177 for all cases except those
that are based on design basis success criteria. As indicated in
WCAP-15049, design basis accumulator success criteria are not
considered necessary to mitigate large break loss-of-coolant
accident (LOCA) events, and were only included in the WCAP-15049
evaluation as a worst case data point. In addition, WCAP-15049
states that the NRC has indicated that an incremental conditional
core damage frequency (ICCDP) greater than 5E-07 does not
necessarily mean the change is unacceptable. The proposed technical
specification change does not involve any hardware changes nor does
it affect the probability of any event initiators. There will be no
change to normal plant operating parameters, engineered safety
feature (ESF) actuation setpoints, accident mitigation capabilities,
accident analysis assumptions or inputs. Therefore, this change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed change. As described in Section 9.1 of the WCAP-
15049 evaluation, the plant design will not be changed with this
proposed technical specification CT increase. All safety systems
still function in the same manner and there is no additional
reliance on additional systems or procedures. The proposed
accumulator CT increase has a very small impact on core damage
frequency. The WCAP-15049 evaluation demonstrates that the small
increase in risk due to increasing the accumulator allowed outage
time (AOT) is within the acceptance criteria provided in RGs 1.174
and 1.177. No new accidents or transients can be introduced with the
requested change and the likelihood of an accident or transient is
not impacted. The malfunction of safety related equipment, assumed
to be operable in the accident analyses, would not be caused as a
result of the proposed technical specification change. No new
failure mode has been created and no new equipment performance
burdens are imposed. Therefore, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change does not involve a significant reduction in
a margin of safety. There will be no change to the departure from
nucleate boiling ratio (DNBR) correlation limit, the design DNBR
limits, or the safety analysis DNBR limits. The basis for the
accumulator LCO, as discussed in Bases Section 3.5.1, is to ensure
that a sufficient volume of borated water will be immediately forced
into the core through each of the cold legs in the event the RCS
pressure falls below the pressure of the accumulators, thereby
providing the initial cooling mechanism during large RCS pipe
ruptures. As described in Section 9.2 of the WCAP-15049 evaluation,
the proposed change will allow plant operation in a configuration
outside the design basis for up to 24 hours, instead of 1 hour,
before being required to begin shutdown. The impact of this on plant
risk was evaluated and found to be very small. That is, increasing
the time the accumulators will be unavailable to respond to a large
LOCA event, assuming accumulators are needed to mitigate the design
basis event, has a very small impact on plant risk. Since the
frequency of a design basis large LOCA (a
[[Page 53112]]
large LOCA with loss of offsite power) would be significantly lower
than the large LOCA frequency of the WCAP-15049 evaluation, the
impact of increasing the accumulator CT from 1 hour to 24 hours on
plant risk due to a design basis large LOCA would be significantly
less than the plant risk increase presented in the WCAP-15049
evaluation. Therefore, this change does not involve a significant
reduction in a margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of amendment request: July 6, 2004.
Description of amendment request: The proposed amendment relocates
the surveillance requirements for Item 22, ``Accumulator Level and
Pressure,'' and Item 25, ``Portable Radiation Survey Instruments,''
from Table TS 4.1-1 of the Technical Specifications to licensee-
controlled documents.
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 amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
NMC [Nuclear Management Company] Response for Proposed Change to
Table TS 4.1-1, Item 22
No. This TS change removes the accumulator water level and
pressure channel surveillance from the TS and places them into
licensee controlled documents. This change is consistent with
industry and NRC [Nuclear Regulatory Commission] recognition that
the accumulator instrumentation operability is not directly related
to the capability of the accumulators to perform their safety
function.
Relocating the instrumentation surveillance requirements is an
administrative change that will not affect equipment testing,
availability, or operation. Therefore, the change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
NMC Response for Proposed Change to Table TS 4.1-1, Item 25
No. Removing the surveillance requirements for portable
radiation survey instruments from the TS is administrative and has
no impact on plant equipment, accident initiators, or the safety
analysis. Additionally, eliminating the monthly check and modifying
the line item description does not impact plant equipment or
operation. Therefore, the change does not involve an increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
NMC Response for Proposed Change to Table TS 4.1-1, Item 22
No. Relocating the accumulator water level and pressure
instrument surveillance requirements to licensee controlled
documents is an administrative change that will not change any
equipment, require new equipment to be installed, or change the way
current equipment operates in the plant.
Therefore, the change will not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
NMC Response for Proposed Change to Table TS 4.1-1, Item 25
No. Removing the surveillance requirements for portable
radiation survey instruments from the TS and relocating the
requirements to licensee controlled documents is administrative and
has no impact on plant equipment or the way the plant equipment
operates. Additionally, eliminating the monthly check and modifying
the line item description does not impact plant equipment or
operation. Portable radiation survey instruments are not accident
initiators. Therefore, the change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
NMC Response for Proposed Change to Table TS 4.1-1, Item 22
No. Relocating the accumulator water level and pressure
instrument surveillance requirements to licensee controlled
documents is an administrative change that will not change the
safety analyses performed for the plant nor reduce the ability of
the accumulators to perform their safety related function. There is
no change in the operation of the accumulators or related equipment
and systems. Therefore, the change does not involve a reduction in
the margin of safety.
NMC Response for Proposed Change to Table TS 4.1-1, Item 25
No. Portable radiation survey instruments are not inputs to the
safety analysis or to automatic plant actions. The change is
administrative since it moves the requirements out of TS and into
licensee controlled documents through use of the 10 CFR 50.36
selection criteria for TS. Additionally, eliminating the monthly
check and modifying the line item description does not impact plant
equipment or operation. Therefore, the change does not reduce the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Section Chief: L. Raghavan.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: August 2, 2004.
Description of amendment request: The proposed amendment would
implement a risk-informed process for determining allowed outage times
for South Texas Project (STP), Units 1 and 2, Technical Specifications
(TS). The risk-informed process involves the application of the STP,
Units 1 and 2, Configuration Risk Management Program (CRMP). The STP
CRMP is a procedurally controlled program utilized for the
implementation of 50.65(a)(4) of Title 10 of the Code of Federal
Regulations (10 CFR).
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 to the Technical Specifications
involve a significant increase in the probability or consequences of
an accident previously evaluated?
The proposed changes to the Technical Specifications to add a
new TS 3.13.1 and to change specific TS to apply the new TS 3.13.1
do not involve a significant increase in the probability of an
accident previously evaluated because the changes involve no change
to the plant or its modes of operation. In addition, the risk-
informed configuration management program will be applied to
effectively manage the availability of required systems, structures,
and components to assure there is no significant increase in the
probability of an accident. These proposed changes do not increase
the consequences of an accident because the design-basis mitigation
function of the affected systems is not changed and the risk-
informed configuration management program will be applied to
effectively manage the availability of systems, structures and
components required to mitigate the consequences of an accident. The
application of the risk-informed configuration management program is
considered a substantial technological improvement over current
methods.
Therefore, none of the proposed changes involve a significant
increase in the
[[Page 53113]]
probability or consequences of an accident previously evaluated.
2. Does the proposed change to the Technical Specifications
create the possibility of a new or different kind of accident from
any accident previously evaluated?
None of the proposed changes involve a new mode of operation or
design configuration. There are no new or different systems,
structures, or components proposed by these changes. Therefore,
there is no possibility of a new or different kind of accident.
3. Does the proposed change to the Technical Specifications
involve a significant reduction to a margin of safety?
Proposed new TS 3.13.1 and the associated changes to the
specifications that apply the new TS 3.13.1 implement a risk-
informed configuration management program to assure that adequate
margins of safety are maintained. Application of these new
specifications and the configuration management program considers
cumulative effects of multiple systems or components being out of
service and does so more effectively than the current Technical
Specifications. Therefore, application of these new specifications
will not involve a significant reduction in a margin of safety.
Based on the evaluation above, none of the proposed changes
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: Robert A. Gramm.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: August 12, 2004.
Description of amendment request: The proposed changes to the South
Texas Project (STP), Units 1 and 2, Technical Specifications (TS) for
steam generators (SGs) are based on draft TS Task Force (TSTF) Improved
Standard TS Change Traveler TSTF-449, Rev. 2, and the Joseph M. Farley
Nuclear Plant, Units 1 and 2, submittal dated June 28, 2004, as
supplemented by letter dated August 5, 2004. The changes would
implement guidance for the industry initiative on Nuclear Energy
Institute (NEI) 97-06, ``Steam Generator Program Guidelines.''
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 requires a Steam Generator Program that
includes performance criteria that will provide reasonable assurance
that the SG tubing will retain integrity over the full range of
operating conditions (including startup, operation in the power
range, hot standby, cooldown, and all anticipated transients
included in the design specification). The SG performance criteria
are based on tube structural integrity, accident induced leakage,
and operational leakage.
The structural integrity performance criterion is:
All inservice SG tubes shall retain structural integrity over
the full range of normal operating conditions (including startup,
operation in the power range, hot standby, and cooldown, and all
anticipated transients included in the design specification) and
design basis accidents. This includes retaining a safety factor of
3.0 (3 [delta] P) against burst under normal steady state full power
operation primary-to-secondary pressure differential and a safety
factor of 1.4 against burst applied to the design basis accident
primary-to-secondary pressure differentials. Apart from the above
requirements, additional loading conditions associated with the
design basis accidents, or combination of accidents in accordance
with the design and licensing basis, shall also be evaluated to
determine if the associated loads contribute significantly to burst
or collapse. In the assessment of tube integrity, those loads that
do significantly affect burst or collapse shall be determined and
assessed in combination with the loads due to pressure with a safety
factor of 1.2 on the combined primary loads and 1.0 on axial
secondary loads.
The accident induced leakage performance criterion is:
The primary-to-secondary accident induced leakage rate for any
design basis accidents, other than a SG tube rupture, shall not
exceed the leakage rate assumed in the accident analysis in terms of
total leakage rate for all SGs and leakage rate for an individual
SG. Accident induced leakage is not to exceed 1 gpm [gallons per
minute] total for all four SGs in a unit.
The operational leakage performance criterion is:
``The RCS operational primary-to-secondary leakage through any
one SG shall be limited to 150 gallons per day.''
An SGTR [steam generator tube rupture] event is one of the
design basis accidents analyzed as part of the plant licensing
basis. In the analysis of an SGTR event, a bounding primary-to-
secondary leakage rate equal to the operational leakage rate limits
in the licensing basis plus the leakage rate associated with a
double-ended rupture of a single tube is assumed.
For other design basis accidents such as MSLB [main steamline
break], rod ejection, and reactor coolant pump locked rotor, the
tubes are assumed to retain their structural integrity (i.e., they
are assumed not to rupture). At STP these analyses assume that the
total primary-to-secondary leakage is 1 gpm. The accident induced
leakage criterion introduced by the proposed changes accounts for
tubes that may leak during design basis accidents. The accident
induced leakage criterion limits this leakage to no more than the
value assumed in the accident analysis.
The SG performance criteria proposed in this change to the TS
identify the standards against which tube integrity is to be
measured. Meeting the performance criteria provides reasonable
assurance that the SG tubing will remain capable of fulfilling its
specific safety function of maintaining RCPB [reactor coolant
pressure boundary] integrity throughout each operating cycle and in
the unlikely event of a design basis accident. The performance
criteria are only a part of the Steam Generator Program required by
the proposed change to the TS. The program, defined by NEI 97-06,
includes a framework that incorporates a balance of prevention,
inspection, evaluation, repair, and leakage monitoring.
The consequences of design basis accidents are, in part,
functions of the dose equivalent I-131 in the primary coolant and
the primary-to-secondary leakage rates resulting from an accident.
Therefore, limits are included in the TS for operational leakage and
for dose equivalent I-131 in primary coolant to ensure the plant is
operated within its analyzed condition. The analysis of the limiting
design basis accident assumes that primary-to-secondary leak rate
after the accident is 1 gpm with no more than 500 gpd [gallons per
day] in any one SG, and that the reactor coolant activity levels of
dose equivalent I-131 are at the TS values before the accident.
The proposed change does not affect the design of the SGs, their
method of operation, or primary coolant chemistry controls. The
proposed approach updates the current TS and enhances the
requirements for SG inspections. The proposed change does not
adversely impact any other previously evaluated design basis
accident and is an improvement over the current TS.
Therefore, the proposed change does not affect the consequences
of an SGTR accident and the probability of such an accident is
reduced. In addition, the proposed changes do not affect the
consequences of an MSLB, rod ejection, or a reactor coolant pump
locked rotor event.
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 performance-based requirements are an improvement
over the requirements imposed by the current TS.
Implementation of the proposed Steam Generator Program will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
result of the implementation of the Steam Generator Program will be
an enhancement of SG tube performance. Primary-to-secondary leakage
[[Page 53114]]
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The SG tubes are an integral part of the RCPB and, as such, are
relied upon to maintain the primary system pressure and inventory.
As part of the RCPB, the SG tubes are unique in that they are also
relied upon as a heat transfer surface between the primary and
secondary systems such that residual heat can be removed from the
primary system. In addition, the SG tubes also isolate the
radioactive fission products in the primary coolant from the
secondary system. In summary, the safety function of a SG is
maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in tube
integrity by implementing the Steam Generator Program to manage SG
tube inspection, assessment, repair, and plugging. The requirements
established by the Steam Generator Program are consistent with those
in the applicable design codes and standards and are an improvement
over the requirements in the current TS.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TS.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: Robert A. Gramm.
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: August 5, 2004.
Brief description of amendments: The proposed change revises
Technical Specification 3.7.10 entitled, ``Control Room Emergency
Filtration/Pressurization System (CREFS),'' to add a new condition for
an inoperable Control Room boundary.
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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
This is a revision to the Technical Specifications for the
Control Room Emergency/Filtration System which is a mitigation
system designed to minimize in leakage and to filter the control
room atmosphere to protect the operator following accidents
previously analyzed. An important part of the system is the Control
Room boundary. The Control Room boundary integrity is not an
initiator or precursor to any accident previously evaluated.
Therefore, the probability of any accident previously evaluated is
not increased. The analysis of the consequences of analyzed accident
scenarios under the control room breach conditions along with the
compensatory actions for restoration of control room integrity
demonstrate that the consequences of any accident previously
evaluated are not increased. Therefore, it is concluded that this
change does not significantly increase the probability [or
consequences] of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change will not impact the accident analysis. The
change will not alter the requirements of the Control Room
Emergency/Filtration System or its function during accident
conditions. The administrative controls and compensatory actions
will ensure the control room emergency/filtration system will
perform its safety function. No new or different accidents result
from performing the new actions and surveillance required. The
change does not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed) or a change in
the methods governing normal plant operation. The change does not
alter assumptions made in the safety analysis. The proposed change
is consistent with the safety analysis assumptions and current plant
operating practice. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not affected by these changes. The proposed change will not
result in plant operation in a configuration outside the design
basis for an unacceptable period of time without compensatory
actions and administrative controls. The proposed change does not
affect systems that respond to safely shutdown the plant and to
maintain the plant in a safe shutdown condition. Therefore the
proposed change does not involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: July 1, 2004.
Description of amendment request: The proposed license amendments
would modify the Reactor Coolant System (RCS) pressure/temperature (P/
T) limit curves, the Low-Temperature Overpressure Protection System
(LTOPS) setpoint allowable values, and the LTOPS Tenable values. In
addition, the cumulative core burnup applicability limits for the LTOPS
would be extended.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes modify the North Anna Units 1 and 2 RCS P/T
limit curves, LTOPS setpoint allowable values, LTOPS Tenable and
extend the cumulative core burnup applicability limits for the
LTOPS. The allowable operating pressures and temperatures under the
proposed RCS P/T limit curves are not significantly different from
those allowed under the existing Technical Specification P/T limits.
The revisions in the values for the LTOPS setpoint allowable values
and LTOPS Tenable values do not significantly change the plant
operating space. No changes to plant systems, structures or
components are proposed, and no new operating modes are established.
The P/T limits, LTOPS setpoint allowable values, and Tenable values
do not contribute to the probability of occurrence or consequences
of accidents previously analyzed. The revised licensing basis
[[Page 53115]]
analyses utilize acceptable analytical methods, and continue to
demonstrate that established accident analysis acceptance criteria
are met. Therefore, there is no increase in the probability or
consequences of any accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes modify the North Anna Units 1 and 2 RCS P/T
limit curves, LTOPS setpoint allowable values, LTOPS Tenable values
and extend the cumulative core burnup applicability limits for the
LTOPS. The allowable operating pressures and temperatures under the
proposed RCS P/T limit curves are not significantly different from
those allowed under the existing Technical Specification P/T limits.
No changes to plant systems, structures or components are proposed,
and no new operating modes are established. Therefore, the proposed
changes do not create the possibility of any accident or malfunction
of a different type previously evaluated.
3. Does the change involve a significant reduction in the margin
of safety?
The proposed revised RCS P/T limit curves, LTOPS setpoint
allowable values, and LTOPS Tenable analysis bases do not involve a
significant reduction in the margin of safety for these parameters.
The effects of RCS pressure and temperature measurement uncertainty
continue to be considered in the supporting analyses. The proposed
revised RCS P/T limit curves are valid to cumulative core burnups of
50.3 EFPY [effective full-power year] and 52.3 EFPY for North Anna
Units 1 and 2 respectively. The proposed revised LTOPS setpoint
allowable values and Tenable analyses support these same cumulative
core burnup limits. The analyses demonstrate that established
analysis acceptance criteria continue to be met. Specifically, the
proposed P/T limit curves, LTOPS setpoint allowable values and LTOPS
Tenable values provide acceptable margin to vessel fracture under
both normal operation and LTOPS design basis (mass addition and heat
addition) accident conditions. Therefore, the proposed changes do
not result in a significant reduction in margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Mary Jane Ross-Lee (Acting).
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: July 22, 2004.
Description of amendment request: The proposed change would revise
Technical Specification (TS) Figure 3.5.5-1, ``Seal Injection Flow
Limits,'' to reflect flow limits that allow a higher seal injection
flow for a given differential pressure between the charging discharge
header and the reactor coolant system pressure. Specifically, the
licensee requests approval of the proposed amendment to allow for
repositioning the seal injection throttle valves during the upcoming
refueling outage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The restriction on reactor coolant pump (RCP) seal injection
flow limits the amount of Emergency Core Cooling System (ECCS) flow
that would be diverted from the injection path following an
accident. This limit is based on safety analysis assumptions that
are required because RCP seal injection flow is not isolated during
safety injection. The intent of the Limiting Condition for Operation
(LCO) limit on seal injection flow is to make sure that flow through
the RCP seal water injection line is low enough to ensure sufficient
centrifugal charging pump injection flow is directed to the Reactor
Coolant System (RCS) via the injection points.
There are no hardware changes nor are there any changes in the
method by which any safety related plant system performs its safety
function. The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility or the manner in which
[the] plant is operated and maintained. The proposed change does not
alter or prevent the ability of structures, systems, and components
from performing their intended safety function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change does not affect the source term,
containment isolation, or radiological release assumptions used in
evaluating the radiological consequences of an accident previously
evaluated. Further, the proposed change does not increase the types
or amounts of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures. The proposed change is consistent with the
safety analysis assumptions and resultant consequences.
Since the change continues to ensure 100 percent of the assumed
charging flow is available, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously analyzed.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There are no hardware changes nor are there any changes in the
method by which any safety related plant system performs its safety
function. This amendment will not affect the normal method of plant
operation. The proposed change does not introduce any new equipment
into the plant or alter the manner in which existing equipment will
be operated. No performance requirements or response time limits
will be affected. The change is consistent with assumptions made in
the safety analysis and licensing basis regarding limits on RCP seal
injection flow.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of this amendment. The[re] will be no adverse effect or challenges
imposed on any safety related system as a result of this amendment.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change does not affect the acceptance criteria for
any analyzed event. There will be no effect on the manner in which
safety limits or limiting safety system settings are determined nor
will there be any effect on those plant systems necessary to assure
the accomplishment of protection function. Increasing the total seal
injection flow limit to 90 gpm does not significantly impact the
assumed ECCS flow that would be available for injection into the RCS
following an accident.
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: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: July 23, 2004.
Description of amendment request: The proposed amendment will
delete the requirements from the technical
[[Page 53116]]
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 proposed license amendment will revise TS 3.3.3, ``Post
Accident Monitoring (PAM) Instrumentation,'' to delete the Note in
Condition C. Also in TS 3.3.3, Condition D will be deleted. In TS Table
3.3.3-1, Function 10, ``Containment Hydrogen Concentration Level,'' is
deleted and replaced with ``Not Used.'' TS 3.6.8, ``Hydrogen
Recombiners,'' will be deleted and the Table of Contents will be
revised to reflect that deletion. TS 5.6.8, ``PAM Report,'' will be
revised to reflect changing Condition G to Condition F in TS 3.3.3.
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 July 23, 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.
RG 1.97 Category 1, is intended for key variables that most directly
indicate the accomplishment of a safety function for design-basis
accident events. The hydrogen 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 SAMGs [severe accident
management guidelines], 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.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: July 23, 2004.
Description of amendment request: The requested change will delete
Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure
Report,'' and TS 5.6.4, ``Monthly Operating Reports.'' The Table of
[[Page 53117]]
Contents will also be revised to reflect the deletions.
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 July 23, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating letter
report of shutdown experience and operating statistics if the
equivalent data is submitted using an industry electronic database.
It also eliminates the TS reporting requirement for an annual
occupational radiation exposure report, which provides information
beyond that specified in NRC regulations. The proposed change
involves no changes to plant systems or accident analyses. As such,
the change is administrative in nature and does not affect
initiators of analyzed events or assumed mitigation of accidents or
transients. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by email to [email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: March 23, 2004.
Brief description of amendment: The amendment eliminates the
Technical Specification requirements related to hydrogen monitors.
Date of Issuance: August 9, 2004.
Effective date: August 9, 2004 and shall be implemented within 60
days of issuance.
Amendment No.: 246.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 27, 2004 (69 FR
22879).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated August 9, 2004.
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
DukeEnergy Corporation, Docket Nos.50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: March 23, 2004.
Brief description of amendments: The amendments revise the reactor
coolant pump flywheel inspection interval from 10 years to 20 years.
Date of issuance: August 5, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 216 and 210, 223 and 205.
Renewed facility operating license Nos. NPF-35, NPF-52, NPF-9, And
NPF-17: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: May 25, 2004.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 5, 2004.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: August 22, 2002, as supplemented
by letters dated September 12, 2003, and February 4, February 16, March
23, April 28, June 17, July 6, July 12, July 19, and July 29, 2004.
[[Page 53118]]
Brief description of amendments: The amendments revised Technical
Specification 3.8.1, ``AC Sources--Operating,'' to temporarily extend
the Completion Times (CTs) for the Keowee hydro units (KHUs) to allow
additional time for maintenance and upgrades. The amendments extend by
17 days (from 45 days to 62 days) the CT when one KHU is not operable
and extend by 120 hours (from 60 hours to 180 hours) the CT when both
KHUs are not operable.
Date of Issuance: August 5, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 339, 341, and 340.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 17, 2002 (67
FR 58641).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 5, 2004.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: October 21, 2003.
Brief description of amendment: The change removes MODE
restrictions that prevent performance of Surveillance Requirements
(SRs) 3.8.4.7 and 3.8.4.8 for the Division III direct current
electrical power subsystem while in MODES 1, 2, or 3. These
surveillances verify that the battery capacity is adequate to perform
its required functions. The changes allow the performance of SR 3.8.4.7
and SR 3.8.4.8 during normal plant operations rather than only during
refueling outages.
Date of issuance: August 12, 2004.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 141.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specfications.
Date of initial notice in Federal Register: December 9, 2003 (68 FR
68662).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 12, 2004.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: March 9, 2004.
Brief description of amendment: The amendment extends the
completion time (CT) from 1 hour to 24 hours for Condition B of
Technical Specification (TS) 3.5.1, which defines requirements for the
emergency core cooling system accumulators. Condition B of TS 3.5.1
specifies a CT to restore an accumulator to operable status when it has
been declared inoperable for a reason other than the boron
concentration of the water in the accumulator not being within the
required range.
Date of issuance: August 18, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 222.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19567).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 18, 2004.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: August 16, 2002, as supplemented
March 25, 2003, April 6, and July 22, 2004.
Brief description of amendment: This amendment deleted the existing
requirements in Technical Specification (TS) 3.10.D.1.d from TS 3/
4.10.D, ``Multiple Control Rod Removal,'' and the associated
Surveillance Requirement 4.10.D.1.d. This amendment added a new
requirement to TS 3.10.D.1.d. Additionally, this amendment made an
editorial change to correct a reference to TS 3.3.B.3 instead of TS
3.3.B.4 in TS 3/4.10.D.1.
Date of issuance: August 17, 2004.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 207.
Facility Operating License No. DPR-35: Amendment revised the TSs.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75873).
The supplements dated March 25, 2003, April 6, and July 22, 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.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 17, 2004.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: February 9, 2004.
Brief description of amendment: The amendment eliminates the
requirements in the Technical Specifications associated with hydrogen
recombiners and hydrogen monitors.
Date of issuance: August 12, 2004.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment No.: 222.
Renewed Facility Operating License No. DPR-51: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: March 30, 2004 (69 FR
16617).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 12, 2004.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: February 9, 2004.
Brief description of amendment: The amendment eliminates the
requirements in the Technical Specifications associated with hydrogen
recombiners and hydrogen monitors.
Date of issuance: August 5, 2004.
Effective date: As of the date of issuance to be implemented within
120 days from the date of issuance.
Amendment No.: 254.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 30, 2004 (69 FR
16618).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 5, 2004.
No significant hazards consideration comments received: No.
[[Page 53119]]
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: August 27, 2003.
Brief description of amendments: The amendments change Technical
Specification 4.0.3, ``Missed Surveillance Time Allowance.'' TS 4.0.3
describes the relationship between meeting the surveillance requirement
and operability. The amendments modify TS 4.0.3 to allow a missed
surveillance to be completed within 24 hours or up to the limit of the
specified interval, whichever is greater. Additionally, the amendments
add a statement that a risk evaluation shall be performed for any
surveillance delayed greater than 24 hours and that the risk impact
shall be managed. The amendments also change the Bases to further
clarify the provisions of the TS. In addition, the proposed amendments
make format changes to improve appearance. The changes to the TS and
its Bases are consistent with industry/Technical Specification Task
Force TSTF-358, Revision 6, which was approved by the Nuclear
Regulatory Commission (NRC) on October 3, 2001, and incorporated the
NRC's comments on TSTF-358, Revision 5. TSTF-358, Revision 5, was
approved with comment by the NRC as a part of the Consolidated Line
Item Improvement Process in a Federal Register Notice dated September
28, 2001.
Date of issuance: August 9, 2004.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 282, 266.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 11, 2004 (69 FR
26190).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 9, 2004.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: January 30, 2004.
Brief description of amendments: The amendments relocate the
requirements for hydrogen monitors to the Technical Requirements
Manual.
Date of issuance: August 13, 2004.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment Nos.: 214 and 219.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 2, 2004 (69 FR
9862).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 13, 2004.
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: March 25, 2004, as supplemented
June 2, 2004.
Brief description of amendments: The amendments approve a change to
the licensing basis to allow the use of the methods described in
Framatome-ANP Topical Report BAW-10169-A, ``RSG Plant Safety Analysis--
B&W Safety Analysis Methodology for Recirculating Steam Generator
Plants,'' dated October 1989, for calculating the mass and energy
release rates resulting from a postulated main steamline break accident
for input to containment analyses. These methods utilize the RELAP5/
MOD2-B&W code approved by the Nuclear Regulatory Commission staff in a
safety evaluation report dated March 14, 1995.
Date of issuance: August 19, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 164 and 155.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
authorized revision to the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: April 27, 2004 (69 FR
22881).
The June 2, 2004, supplemental letter contained clarifying
information and did not change the initial proposed no significant
hazards consideration determination and was within the scope of the
original Federal Register notice.
The Commission's related evaluation of the amendments is contained
in a safety evaluation dated August 19, 2004.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of application for amendment: July 23, 2003.
Brief description of amendment: Revised the near end-of-life
Moderator Temperature Coefficient (MTC) Surveillance Requirement
4.1.1.3.b by placing a set of conditions on core operation, which if
met, would allow exemption from the required MTC measurement. The
conditional exemption is determined on a cycle-specific basis by
considering the margin predicted to the surveillance requirement MTC
limit and the performance of other core parameters, such as beginning
of life MTC measurements and the critical boron concentration as a
function of cycle life.
Date of issuance: July 21, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 169.
Renewed Facility Operating License No. NPF-12: Amendment revises
the Technical Specifications.
Date of initial notice in Federal Register: September 30, 2003 (68
FR 56346).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 21, 2004.
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 Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: December 30, 2003.
Brief description of amendments: The amendments revised the staff
position titles in Section 5.0 ``Administrative Controls'' of the
Technical Specifications.
Date of issuance: June 3, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 242 and 185.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 2, 2004 (69 FR
9865).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 3, 2004.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 23rd day of August 2004.
[[Page 53120]]
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Division of Licensing Project Management, Office of Nuclear Reactor
Regulation.
Director,
[FR Doc. 04-19586 Filed 8-30-04; 8:45 am]
BILLING CODE 7590-01-P