[Federal Register Volume 69, Number 187 (Tuesday, September 28, 2004)]
[Notices]
[Pages 57978-57999]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-21345]
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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 September 3, 2004, through September 16,
2004. The last biweekly notice was published on September 14, 2004 (69
FR 55466).
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
[[Page 57979]]
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
O1F21, 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 basis for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the
[[Page 57980]]
Office of the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff; (2) courier, express mail, and expedited delivery
services: Office of the Secretary, Sixteenth Floor, One White Flint
North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention:
Rulemaking and Adjudications Staff; (3) e-mail addressed to the Office
of the Secretary, U.S. Nuclear Regulatory Commission,
[email protected]; or (4) facsimile transmission addressed to the
Office of the Secretary, U.S. Nuclear Regulatory Commission,
Washington, DC, Attention: Rulemakings and Adjudications Staff at (301)
415-1101, verification number is (301) 415-1966. A copy of the request
for hearing and petition for leave to intervene should also be sent to
the Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and it is requested that copies be
transmitted either by means of facsimile transmission to (301) 415-3725
or by e-mail to [email protected]. A copy of the request for
hearing and petition for leave to intervene should also be sent to the
attorney for the licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer 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 O1F21, 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 e-mail
to [email protected].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendments request: July 9, 2004.
Description of amendments request: The amendments would revise the
Technical Specifications (TSs) to allow operation of Palo Verde Nuclear
Generating Station (PVNGS), Units 1 and 3 up to a maximum reactor core
power level of 3990 Megawatts thermal (MWt), an increase of 2.94
percent above the current licensed power level of 3876 MWt. The
proposed amendments would also make administrative changes to the PVNGS
Unit 2 TSs so that the changed pages would apply to the three PVNGS
units. Operation at the uprated power level with replacement steam
generators has been approved for PVNGS Unit 2.
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.
(a) Evaluation of the Probability of Previously Evaluated
Accidents
Plant Structures, Systems and Components (SSCs) have been
verified to be capable of performing their intended design functions
at uprated power conditions. Where necessary, a small number of
minor modifications will be made prior to implementation of uprated
power operations so that surveillance test acceptance criteria
continues to be met. The analysis has concluded that operation at
uprated power conditions will not adversely affect the capability or
reliability of plant equipment. Current technical specification (TS)
surveillance requirements ensure frequent and adequate monitoring of
system and component operability. All systems will continue to be
operated within current operating requirements at uprated
conditions. Therefore, no new structure, system or component
interactions have been identified that could lead to an increase in
the probability of any accident previously evaluated in the Updated
Final Safety Analysis Report (UFSAR).
(b) Evaluation of the Consequences of Previously Evaluated
Accidents
The radiological consequences were reviewed for all design basis
accidents (DBAs) (i.e., both LOCA [loss-of-coolant accident] and
non-LOCA accidents) previously analyzed in the UFSAR. The analysis
showed that the resultant radiological consequences for both LOCA
and non-LOCA accidents remain either unchanged or have not
significantly increased due to operation at uprated power
conditions. The radiological consequences of all DBAs continue to
meet established regulatory limits.
(2) Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated.
Response: No.
The configuration, operation and accident response of the PVNGS
[Palo Verde Nuclear Generating Station] Units I and 3 structures,
systems, and components are unchanged by operation at uprated power
conditions or by the associated proposed TS changes. Analyses of
transient events have confirmed that no transient event results in a
new sequence of events that could lead to a new accident or
different scenario.
The effect of operation at uprated power conditions on plant
equipment has been evaluated. No new operating mode, safety-related
equipment lineup, accident scenario, or equipment failure mode was
identified as a result of operating at uprated conditions. In
addition, operation at uprated power conditions does not create any
new failure modes that could lead to a different kind of accident.
Minor plant modifications, to support Implementation of uprated
power conditions, will be made as required to existing SSCs. The
basic design function of all SSCs remains unchanged and no new
equipment or systems have been installed that could potentially
introduce new failure modes or accident sequences.
Based on this analysis, it is concluded that no new accident
scenarios, failure mechanisms or limiting single failures are
introduced as a result of the proposed changes. The proposed changes
do not have an adverse effect on any safety-related system or design
basis function. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) Does the proposed changes involve a significant reduction in
a margin of safety?
Response: No.
A comprehensive analysis was performed to evaluate the effects
of power uprate on PVNGS Units 1 and 3. This analysis identified and
defined the major input parameters to the NSSS [nuclear steam supply
system], reviewed NSSS design transients, and reviewed the
capabilities of the NSSS and BOP [balance of plant] fluid systems,
NSSS/BOP interfaces, NSSS and BOP control systems, and NSSS and BOP
SSCs. All appropriate NSSS accident analyses were re-performed to
confirm that acceptable results were maintained and that the
radiological consequences remained within regulatory and Standard
Review Plan (SRP) limits. The nuclear and thermal hydraulic
performance of nuclear fuel was also reviewed to confirm acceptable
results. The analyses confirmed that all NSSS and BOP SSCs are
capable, some with minor modifications, to safely support operations
at uprated power conditions.
The margin of safety of the reactor coolant pressure boundary is
maintained under uprated power conditions. The design pressure of
the reactor pressure vessel and reactor coolant system will not be
challenged as the pressure mitigating systems were confirmed to be
sufficiently sized to adequately control pressure under uprated
power conditions.
Reanalysis of containment structural integrity under Design
Basis Accident (DBA) conditions indicates that the calculated peak
[[Page 57981]]
containment pressure (Pa) increases from 52.0 psig [pounds per
square inch gauge] to 58.0 psig, but remains less than the
containment internal design pressure of 60 psig. The proposed value
for Pa has been rounded up from the actual calculated value of 57.85
psig.
Radiological consequences of the following accidents were
reviewed: Main Steam Line Break, Locked Reactor Coolant Pump (RCP)
Rotor, CEA Ejection, Small Steam Line Break Outside Containment,
Steam Generator Tube Rupture, LBLOCA, SBLOCA, Waste Gas Decay Tank
Rupture, Liquid Waste Tank Failure, and Fuel Handling Accident. The
resultant radiological consequences for each of these accidents did
not show a significant change due to uprated power conditions and 10
CFR 100 and SRP limits continue to be met.
The analyses supporting operation at power uprate conditions
have demonstrated that all systems and components are capable of
safely operating at uprated power conditions. All design basis
accident acceptance criteria will continue to be met. Therefore, it
is concluded that the 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
that review, it appears that the three 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: Kenneth C. Manne, Senior Attorney, Arizona
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix,
Arizona 85072-2034.
NRC Section Chief: Robert Gramm.
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: August 19, 2004.
Description of amendment request: The proposed amendment would
revise the reactor coolant system (RCS) pressure and temperature limits
by replacing Technical Specification Section 3.4.3, ``RCS Pressure and
Temperature (P/T) Limits,'' Figures 3.4.3-1 and 3.4.3-2, with figures
that are applicable up to 35 effective full-power years (EFPY).
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?
The proposed RCS P/T limits are based on NRC-approved
methodology and will continue to maintain appropriate limits for the
HBRSEP [H.B. Robinson Steam Electric Plant], Unit No. 2, RCS up to
35 EFPY. These changes provide appropriate limits for pressure and
temperature during heatup and cooldown of the RCS, thus ensuring
that the probability of RCS failure is maintained acceptably low.
These limits are not directly related to the consequences of
accidents.
Therefore, the proposed changes do not involve a significant
increase in 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 previously evaluated?
The proposed changes will continue to ensure that the RCS will
be maintained within appropriate pressure and temperature limits
during heatup and cooldown. No physical changes to the HBRSEP, Unit
No. 2, systems, structures, or components are being implemented.
There are no new or different accident initiators or sequences being
created by the proposed Technical Specifications changes. Therefore,
these changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
(3) Do the proposed changes involve a significant reduction in
the margin of safety?
The proposed changes ensure that the margin of safety for the
fission product barriers protected by these functions will continue
to be maintained. This conclusion is based on use of the applicable
NRC-approved methodology for developing and establishing the
proposed RCS P/T limits. Therefore, these changes do not involve a
significant reduction in the margin of safety.
Based on the preceding discussion, 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 Marshall (Acting).
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant, Middlesex County, Connecticut
Date of amendment request: August 11, 2004.
Description of amendment requests: The Haddam Neck Plant (HNP) is
currently undergoing active decommissioning. The proposed amendment
would revise Technical Specifications (TS) to reflect removal of all
Spent Nuclear Fuel (SNF) from the HNP spent fuel pool, and delete the
requirement for submittal of an annual Occupational Radiation Exposure
Report consistent with Industry's Technical Specifications Task Force
(TSTF)-369, Revision 1.
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:
In accordance with 10 CFR 50.92, CYAPCO has reviewed the
proposed changes and concluded that the proposed changes do not
involve a Significant Hazard Consideration (SHC). The following is
provided in support of this conclusion:
Incorporation of TSTF-369, Revision 1: CYAPCO has reviewed the
no significant hazards consideration determination published in the
Federal Register (69 FR 35067) as part of the CLIIP. CYAPCO has
concluded that the determination presented in the Federal Register
is applicable to the HNP and is hereby incorporated by reference to
satisfy the requirements of 10 CFR 50.91.
Deletion and Relocation of Technical Specifications: The
proposed changes do not involve an SHC because the changes would
not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes (deletion of operational requirements and
certain design requirements) reflect the complete transfer of the
spent fuel from the spent fuel pool to the Independent Spent Fuel
Storage Installation (ISFSI). Design basis accidents related to the
spent fuel pool are discussed in the Haddam Neck Plant (HNP) Updated
Final Safety Analysis (UFSAR) Chapter 15. These postulated accidents
are predicated on spent fuel being stored in the spent fuel pool.
With the removal of the spent fuel from the spent fuel pool, there
are no remaining safety related Structures, Systems, and Components
(SSCs) to be monitored and there are no credible accidents that
require the actions of a Certified Fuel Handler or an Equipment
Operator to prevent occurrence or mitigate the consequences of an
accident.
In addition, the HNP UFSAR Chapter 15 also provides a discussion
of other radiological events postulated to occur as a result of
decommissioning with the bounding consequences resulting from a fire
in a resin container. The proposed changes do not have an adverse
impact on decommissioning activities or any of their postulated
consequences.
The proposed changes related to the relocation of certain
administrative requirements do not affect operating procedures or
administrative controls that have the function of preventing or
mitigating any design basis accidents. In addition, these proposed
changes are consistent with the guidance of NRC Administrative
Letter 95-06.
Therefore, the proposed changes do not involve a significant
increase in the
[[Page 57982]]
probability or consequences of any accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed changes eliminate the operational requirements and
certain design requirements associated with the storage of the spent
fuel in the spent fuel pool, and relocate certain administrative
controls to the Connecticut Yankee Quality Assurance Program
(CYQAP). With the complete removal of the spent fuel from the spent
fuel pool, there are no safety related SSCs that remain at the
plant. Thus the proposed changes will not have any effect on the
operation or design function of safety related SSCs. The proposed
changes do not introduce any new failure modes. Therefore, the
proposed changes will not create the possibility of a new or
different kind of accident from any previously evaluated.
(3) Involve a significant reduction in a margin of safety.
The design basis and accident assumptions within the HNP UFSAR
and the Technical Specifications relating to spent fuel are no
longer applicable. The proposed changes do not affect remaining
plant operations, systems, or components supporting decommissioning
activities. In addition, the proposed changes do not result in a
change in initial conditions, system response time, or in any other
parameter affecting the course of a decommissioning activity
accident analysis. Therefore, the proposed changes 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.
NRC Section Chief: Claudia Craig.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: May 27, 2004.
Description of amendment request: The proposed amendment would
delete the requirements from the Technical Specifications (TS) to
maintain hydrogen recombiners and hydrogen monitors. A notice of
availability for the TS improvement using the consolidated line item
improvement process was published in the Federal Register on September
25, 2003 (68 FR 554416). Licensees were generally required to implement
upgrades as described in NUREG-0737, ``Clarification of TMI [Three Mile
Island] Action Plan Requirements,'' and Regulatory Guide (RG) 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI, Unit 2. Requirements related to
combustible gas control were imposed by Order for many facilities and
were added to or included in the TSs for nuclear power reactors
currently licensed to operate.
The revised 10 CFR 50.44, ``Standards for Combustible Gas Control
System in Light-Water-Cooled Power Reactors,'' eliminated the
requirements for hydrogen recombiners and relaxed safety
classifications and licensee commitments to certain design and
qualification criteria for hydrogen and oxygen monitors.
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on September
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the
model NSHC determination in its application dated May 27, 2004. Basis
for proposed no significant hazards consideration determination: As
required by 10 CFR 50.91(a), an analysis of the issue of no significant
hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen monitors are no longer required to mitigate design-basis
accidents and, therefore, the hydrogen monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
Category 1 in RG 1.97 is intended for key variables that most
directly indicate the accomplishment of a safety function for
design-basis accident events. The hydrogen monitors no longer meet
the definition of Category 1 in RG 1.97. As part of the rulemaking
to revise 10 CFR 50.44, the Commission found that Category 3, as
defined in RG 1.97, is an appropriate categorization for the
hydrogen monitors because the monitors are required to diagnose the
course of beyond design-basis accidents.
The regulatory requirements for the hydrogen monitors can be
relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3, and removal
of the hydrogen monitors from TS will not prevent an accident
management strategy through the use of the severe accident
management guidelines, the emergency plan, the emergency operating
procedures, and site survey monitoring that support modification of
emergency plan protective action recommendations.
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen monitor requirements,
including removal of these requirements from TSs, does not involve a
significant increase in the probability or the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident from any Previously Evaluated
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TSs, will not result in any failure mode
not previously analyzed. The hydrogen recombiner and hydrogen
monitor equipment was intended to mitigate a design-basis hydrogen
release. The hydrogen recombiner and hydrogen monitor equipment are
not considered accident precursors, nor does their existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within
the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from 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
[[Page 57983]]
approximately 24 hours after the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI, Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Therefore, this change does not involve a significant reduction
in the margin of safety. Removal of hydrogen monitoring from TSs
will not result in a significant reduction in their functionality,
reliability, and availability.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: Mary Jane Ross-Lee (Acting).
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: August 20, 2004.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3.3.8, ``Post Accident Monitoring
[PAM] Instrumentation,'' to eliminate TS requirements associated with
the reactor building spray (RBS) flow instruments commensurate with the
importance of their revised post-accident function.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1--The proposed amendment would not involve a significant
increase in the probability or consequences of an accident previously
evaluated
Duke proposes to remove the RBS flow instrument from Technical
Specification Table 3.3.8-1 based on a change in its purpose due to
recent modifications completed at Oconee. The TS 3.3.8 requirement
to declare the affect [affected] RBS System train inoperable is
conservative (and inappropriate) when the associated RBS flow
instrument is inoperable. Due to recent plant modifications, the RBS
flow instruments are no longer needed to allow the operator to
throttle flow to preclude RBS pump runout post accident. The revised
post accident function of this PAM instrument is to provide
information to indicate the operation of the RBS System. There are
alternate means to verify that the RBS is in operation, such as,
verifying the RBS pump and valve status. The failure of an RBS flow
instrument has no impact on the probability of an accident analyzed
in the UFSAR [Updated Final Safety Analysis Report]. The RBS flow
instrument is no longer needed to mitigate the consequences of an
accident analyzed in the UFSAR. As such, the proposed LAR [license
amendment request] does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The proposed amendment would not create the possibility of
a new or different kind of accident from any kind of accident
previously evaluated
Duke proposes to remove the RBS flow instrument from Technical
Specification Table 3.3.8-1 based on a change in its purpose due to
recent modifications completed at Oconee. The TS 3.3.8 requirement
to declare the affect [affected] RBS System train inoperable is
conservative (and inappropriate) when the associated RBS flow
instrument is inoperable. Due to recent plant modifications, the RBS
flow instruments are no longer needed to allow the operator to
throttle flow to preclude RBS pump runout post accident. These
changes do not alter the nature of events postulated in the Safety
Analysis Report nor do they introduce any unique precursor
mechanisms. Therefore, the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Criterion 3--The proposed amendment would not involve a significant
reduction in a margin of safety
The proposed TS changes do not unfavorably affect any plant
safety limits, set points, or design parameters. The changes also do
not unfavorably affect the fuel, fuel cladding, RCS [reactor coolant
system], or containment integrity. Therefore, the proposed TS
change, which changes TS requirements associated with revised PAM
function of the RBS flow instrument channels, does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anne W. Cottingham, Winston and Strawn LPP,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Mary Jane Ross-Lee, Acting.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: August 26, 2004.
Description of amendment request: The proposed amendments would add
new Technical Specification (TS) 3.3.29 and TS Bases 3.3.29, ``Reactor
Building Auxiliary Cooler (RBAC) Isolation Circuitry,'' to accommodate
new circuitry that isolates non-safety portions of the low pressure
service water (LPSW) system piping inside containment that supply the
RBACs. This isolation eliminates potentially damaging water hammers
that could occur in the event of certain design-bases events or
transients.
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:
Criterion 1--The Proposed Amendment Would Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The requested license amendment would add a new Technical
Specification to provide appropriate controls for the Reactor Building
(RB) Auxiliary Cooler (RBAC) isolation circuitry that is being added to
the design of the three Oconee units. The RBAC isolation circuitry
provides an automatic means to isolate the LPSW flow stream to the
RBACs on a loss of LPSW flow that can lead to a column closure water
hammer inside the RB when LPSW flow is restarted. The new circuitry
ensures that significant waterhammers do not occur in the LPSW piping
to the RBACs and other RB components. The new circuitry will eliminate
an Operable but degraded/non-conforming condition associated with
potentially damaging waterhammers.
The proposed RBAC isolation circuitry Technical Specification will
provide means to assure that the RBAC isolation circuitry operates at a
performance level necessary to provide for safe operation of the LPSW
system following installation of the LPSW modification and RBAC
isolation circuitry at each of the three units. The addition of the
RBAC isolation circuitry Technical Specification does not increase the
probability or consequences of any accident previously evaluated.
Criterion 2--The Proposed Amendment Would Not Create the Possibility of
a New or Different Kind of Accident From Any Accident Previously
Evaluated
The proposed RBAC isolation circuitry Technical Specification
provides a means to assure the isolation circuitry operates at a
performance level necessary to provide for safe operation
[[Page 57984]]
of the modified LPSW system flow to the RBACs. The change enhances the
plant design by eliminating the possibility of significant waterhammers
that could occur inside the RB on a loss of LPSW flow to the RBACs.
The proposed Technical Specification will not create the
possibility of a new or different kind of accident from any kind of
accident previously evaluated.
Criterion 3--The Proposed Amendment Would Not Involve a Significant
Reduction in a Margin of Safety.
The proposed change does not adversely affect any plant safety
limits, set points, or design parameters. The change also does not
adversely affect the fuel, fuel cladding, Reactor Coolant System, or
containment integrity. The RBACs will continue to be isolated during
ES events. The modification eliminates significant waterhammers in
the LPSW piping to the RBACs.
The change will enhance the ability to provide LPSW flow to
safety related loads following LOOP events. 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: Anne W. Cottingham, Winston and Strawn LPP,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Mary Jane Ross-Lee, Acting.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: June 24, 2004.
Description of amendment request: The proposed amendment would
allow 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 TSs, 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 would be
revised to reflect the related changes to LCO 3.0.4, and Surveillance
Requirement (SR) 3.0.4 would be revised to reflect the LCO 3.0.4
allowance.
This change was proposed by the industry's TS 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 June 24, 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 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.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc. 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: September 1, 2004.
Description of amendment request: The proposed amendment would
delete Technical Specification (TS) 5.6.1, ``Occupational Radiation
Exposure Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated September 1, 2004.
Basis for proposed no significant hazards consideration determination:
As required by 10 CFR 50.91(a), an analysis of the issue of no
significant
[[Page 57985]]
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 eliminates the TS reporting requirements to
provide a monthly operating report of shutdown experience and
operating statistics if the equivalent data is submitted using an
industry electronic database. It also eliminates the Technical
Specification 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.
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, 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.
Criterion 3--The proposed change does not involve a significant
reduction in a margin of safety?
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve significance hazards consideration.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Dockets Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of application for amendments: June 24, 2004.
Description of amendment request: The proposed amendment would
incorporate several Technical Specification Task Force (TSTF) changes
to the licensees Technical Specifications (TSs). The specific TSTF
changes that would be incorporated are:
(1) TSTF-5, Rev. 1, Delete Safety Limit Violation Notification
Requirement--This change modifies TS Section 2.2 to remove the
requirements to report safety limit violations. Associated references
to Title 10 of the Code of Federal Regulations (10 CFR), Sections 50.72
and 50.73, are also removed.
TSTF-208, Rev. 0, Extension of Time to Reach Mode 2 in LCO
(Limiting Condition for Operation) 3.0.3--This TSTF modifies TS Section
LCO 3.0.3 to revise the time to be in Mode 2 once LCO 3.0.3 is entered
from 7 hours to a bracketed site-specific time depending on the
individual plant's ability to reach Mode 2 in a controlled shutdown.
TSTF-222, Rev. 1, Control Rod Scram Time Testing and TSTF-229, Rev.
0, Revise Surveillance Requirement 3.2.2.2 for Consistency with
3.1.4.4--This TSTF modifies the TSs to clarify the frequency of
performing control rod scram time testing subsequent to performance of
an outage that involved the movement of fuel. The current wording of
Surveillance Requirement (SR) 3.1.4.1 could be interpreted that all
control rods need to be scram time tested even if the shutdown was for
a brief amount of time and only a limited amount of fuel was moved in
the reactor (e.g., if only one bundle is moved in a mid-cycle fuel
replacement). This change clarifies the intent of the TSs.
TSTF-297, Rev. 1, and TSTF-227, Rev. 0--These two TSTFs affect the
following three TS Sections:
3.3.2.2--Feedwater and Main Turbine High Water Level Trip
Instrumentation
3.3.4.1--Anticipated Transient Without Scram Recirculation Pump Trip
(ATWS-RPT) Instrumentation
3.3.4.2--End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation
TSTF-297, Rev. 1--This TSTF modifies the TSs to add a new Required
Action and corresponding note to allow affected feedwater pump(s) and
main turbine valve(s) to be removed from service. This change is
necessary to allow components to be removed from service to fulfill the
safety function without a reduction in power to less than 25% rated
thermal power. A similar note is added to TS Sections 3.3.4.1 and
3.3.4.2 to provide the same clarification for when the associated
Required Action is the appropriate action.
TSTF-227, Rev. 0--This TSTF modifies the TSs to eliminate ambiguity
in the EOC-RPT Instrumentation Condition A. Since the LCO allows for
having EOC-RPT instrumentation OPERABLE or certain fuel thermal limits
are met, Condition A was inappropriately worded. The wording of
Condition A is revised to add the word `required' if one or more
channels are inoperable. Without the word `required', one could
interpret Condition A as needing entry even if the fuel thermal limits
were being applied instead of applying the operability requirements to
the EOC-RPT instrumentation.
TSTF-295, Rev. 0, Post-Accident Monitoring Clarifications--This
TSTF modifies the TSs to clarify that a separate Condition entry is
allowed for each penetration flow path for the Post Accident Monitoring
(PAM) instrumentation Primary Containment Isolation Valve (PCIV)
indication function.
TSTF-275, Rev. 0, ECCS Instrumentation Clarifications--This TSTF
modifies the TSs to clarify which Emergency Core Cooling System (ECCS)
instrumentation is required to be OPERABLE to support Emergency Diesel
Generator (EDG) operability. Footnote (a) to Table 3.3.5.1-1 has been
changed to only require the affected functions to be OPERABLE in Modes
4 and 5 when the associated ECCS is required to be OPERABLE per LCO
3.5.2.
TSTF-306, Rev. 2, Traversing In-Core Probe Instrumentation
Specification Requirements--This TSTF modifies the TSs by adding a note
that penetration flow path may not be isolated intermittently under
administrative control to conform to what is already allowed for
similar specifications for Primary Containment Isolation Valves
(PCIVs). Also, the Traversing In-core Probe (TIP) system isolation is
set apart as a separate function including the allowance of isolating
the penetration instead of requiring a plant shutdown.
TSTF-416, Rev. 0, Clarification of LPCI Operability during Decay
Heat Removal Operations--This TSTF modifies the TSs by moving the note
that modifies Low Pressure Coolant Injection (LPCI) surveillances to
the LCO in LCO 3.5.1 and LCO 3.5.2. These notes provide clarity that
the LPCI may be considered OPERABLE during alignment and operation in
the decay heat removal Mode.
TSTF-17, Rev. 2, Containment Airlock Testing Frequency--This TSTF
modifies the TSs to extend the testing frequency of the containment
interlock mechanism from 184 days to 24 months. Also, the corresponding
note for this surveillance is no longer required due to the longer
surveillance frequency.
[[Page 57986]]
TSTF-30, Rev. 3, TSTF-323, Rev. 0, TSTF-45, Rev. 2, TSTF-46, Rev.
1, and TSTF-269, Rev. 2, Containment Isolation Valve Specification
Changes--These TSTFs modify TS Sections 3.6.1.3 concerning Primary
Containment Isolation Valves (PCIVs) and 3.6.4.2 concerning Secondary
Containment Isolation Valves (SCIVs).
TSTF-30, Rev. 3 & TSTF-323, Rev. 0--These TSTFs revise TS 3.6.1.3
to allow for a 72-hour completion time for a closed system flow path
with an inoperable isolation valve and allow for a 72-hour completion
time for a penetration flow path with an inoperable Excess Flow Check
Valve (EFCV).
TSTF-45, Rev. 2--This TSTF revises TSs 3.6.1.3 and 3.6.4.2 to
revise surveillance requirements for valve line-ups. Specifically, if a
containment isolation valve is locked, sealed, or otherwise secured,
they are not required to be verified to be closed during the
performance of the surveillance test.
TSTF-46, Rev. 1--This TSTF revises containment isolation valve
surveillances to delete the reference to verifying the isolation time
of `each power operated' containment isolation valve and only require
verification of each `automatic isolation valve'.
TSTF-269, Rev. 2--This TSTF allows for verification of valve status
by administrative means for repetitive verification of locked, sealed,
or secured valves.
TSTF-322, Rev. 2, Secondary Containment Operability Clarification--
This TSTF modifies the TSs to clarify the intent of the secondary
containment boundary integrity. Associated surveillances currently
imply that secondary containment would be inoperable if a Standby Gas
Treatment (SGT) subsystem was inoperable.
TSTF-276, Rev. 2, Power Factor for Emergency Diesel Generator (EDG)
Surveillances--This TSTF modifies the TSs to allow for certain EDG
testing to be performed even if the specified power factor cannot be
achieved.
TSTF-65, Rev. 1, Generic Organization Titles--This TSTF modifies
the TSs to allow the use of generic organizational titles in place of
plant-specific titles. Therefore, for the TSs, a change is requested to
replace plant-specific titles with generic titles.
TSTF-299, Rev. 0, Primary Coolant Sources Inspection Requirements--
This TSTF modifies the TSs Section 5.2.2, `Primary Coolant Sources
Outside Containment' to clarify the intent of refueling cycle intervals
with respect to the system leak test requirements and adds a sentence
that the leak test is subject to the provisions of Surveillance
Requirements (SR) 3.0.2.
TSTF-279, Rev. 0, Inservice Testing Program Clarifications--This
TSTF modifies TSs Section 5.5.8, ``Inservice Testing Program,'' to
delete the reference to `applicable supports' as part of the
description for the Inservice Testing Program. The applicable TS
Section is 5.5.6.
TSTF-118, Rev. 0, Diesel Generator Fuel Oil Testing Program
Clarifications--This TSTF modifies TSs Section 5.5.13, ``Diesel Fuel
Oil Testing Program,'' to allow for the provisions of SR 3.0.2 (25%
extension) and SR 3.0.3 (missed surveillance actions) to apply to
surveillances. The applicable TS Section is 5.5.9.
TSTF-106, Rev. 1, Diesel Generator Fuel Oil Testing Program
Clarifications--This TSTF modifies the TSs to clarify that Section
5.5.10.b, concerning verification of the diesel fuel oil that was
sampled meets the required ASTM properties, only applies to new fuel.
As written, it could be interpreted that this testing is required for
existing fuel that is routinely sampled. The applicable TS Section is
5.5.9.b.
TSTF-152, Rev. 0, Routine Reporting Requirements Upgrade--This TSTF
modifies the TSs to revise the Occupational Radiation Exposure Report
and the Radioactive Effluent Release Report requirements to be
consistent with other regulatory changes that have occurred.
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:
A. TSTF-5, Rev. 1, Delete Safety Limit Violation Notification
Requirements.
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This action does not affect the plant or operation of the plant.
The change simply removes duplicative information from the Technical
Specifications that is covered in the NRC regulations. Therefore,
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
All systems, structures, and components previously required for the
mitigation of a transient remain capable of fulfilling their
intended design functions. The proposed changes have no adverse
effects on any safety-related system or component and do not
challenge the performance or integrity of any safety related system.
This change is considered an administrative action to remove
duplicative reporting requirements.
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 administrative action does not involve any reduction in a
margin of safety. Removal of duplicative information does not affect
compliance with the regulations. Therefore, the proposed change does
not involve a significant reduction in a margin of safety.
B. TSTF-208, Rev. 0, Extension of Time to Reach Mode 2 in LCO
3.0.3.
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The time frame to take response action in accordance with LCO
3.0.3 is not an initiating condition for any accident previously
evaluated and the accident analyses do not assume that any equipment
is out of service such that LCO 3.0.3 is entered. The small increase
in the time allowed to reach Mode 2 would not place the plant in any
significantly increased probability of an accident occurring. The
plant would already be proceeding to a plant shutdown condition
because of the 1 hour requirement to initiate shutdown actions.
There is no change in the time period to reach Mode 3. The Mode 3
Condition is the point where the plant is shutdown. Therefore, since
there is no change to the 1 hour requirement to initiate the
shutdown nor any change to the time period to reach the shutdown
Condition, the small change in the time to reach the Mode 2 status
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
There are no plant physical alterations proposed. The proposed
changes have no adverse effects on any safety-related system or
component and do not challenge the performance or integrity of any
safety related system. Therefore, the proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
The time period to reach Mode 3 and Mode 4 are unaffected by
this activity. This change simply provides a plant specific value
for reaching Mode 2 if LCO 3.0.3 is entered
[[Page 57987]]
which is within the intent of LCO 3.0.3 for performing a controlled
plant shutdown. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
C. TSTF-222. Rev. 1, Control, Red Scram Time Testing, and TSTF-
229, Rev. 0, Revise Surveillance Requirement 3.2.2.2 for Consistency
with 3.1.4.4
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
These changes are considered clarifications to the original
intent of the Technical Specifications. Adequate testing of control
rods is ensured by this change. Control rod operability is not
affected by these changes. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
All systems, structures, and components previously required for the
mitigation of a transient remain capable of fulfilling their
intended design functions. The proposed changes have no adverse
effects on any safety-related system or component and do not
challenge the performance or integrity of any safety-related system.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
This change is administrative in nature and does not affect any
safety analyses assumptions. Adequate control rod testing continues
to be maintained with implementation of this activity. Therefore,
the proposed change does not involve a significant reduction in a
margin of safety.
(D) TSTF 297, Rev. 1, and TSTF 227, Rev. 0, Enhancements to
Feedwater/Main Turbine High Water Level Trip, EOC-RPT, and ATWS RPT
Specifications
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
There are no changes to the plant configuration assumed for any
accident. The removal from service of equipment that results in its
safety function being met can not adversely affect the consequences
of accidents previously evaluated. Other changes are administrative
clarifications that have no affect on accidents. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
All systems, structures, and components previously required for the
mitigation of a transient remain capable of fulfilling their
intended design functions. The proposed changes have no adverse
effects on any safety-related system or component and do not
challenge the performance or integrity of any safety-related system.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
The actions involved with this activity ensure that safety
functions are met. There are no changes in the overall requirements
of having trip instrumentation available for event mitigation. There
are no affects on the plant safety analyses. Therefore, the proposed
change does not involve a significant reduction in a margin of
safety.
(E) STF-295, Rev. 0, Post-Accident Monitoring Clarications
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The equipment involved with the revised Technical Specifications
are for post-accident monitoring. This equipment has no possibility
of increasing the probability of occurrence of the accident since it
is monitoring equipment only. The consequences of an accident are
not affected since this change maintains the original intent of the
Technical Specifications in having available monitoring information
for each PCIV penetration path. Therefore, the proposed change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
All systems, structures, and components previously required for the
mitigation of a transient remain capable of fulfilling their
intended design functions. The proposed changes have no adverse
effects on any safety-related system or component and do not
challenge the performance or integrity of any safety related system.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
The Technical Specifications continue to require appropriate
post accident monitoring equipment to be OPERABLE. Adequate
instrumentation for post-accident monitoring will be ensured by the
Technical Specification requirements. There are no changes to the
plant safety analyses involved with this change. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety.
(F) TSTF-275, Rev. 0, ECCS Instrumentation Clarifications
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The equipment involved is for mitigative purposes and will not
affect the probability of occurrence of an accident. Technical
Specifications ensures that adequate mitigative equipment continues
to be OPERABLE for any event that may occur in Modes 4 and 5. This
change is considered an upgrade to the specifications that will
provide more consistency within the Technical Specifications. There
are no changes to requirements that ensure appropriate Emergency
Core Cooling Systems are OPERABLE. Therefore, the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
All systems, structures, and components previously required for the
mitigation of a transient remain capable of fulfilling their
intended design functions. The proposed changes have no adverse
effects on any safety-related system or component and do not
challenge the performance or integrity of any safety-related system.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
There is no impact on mitigative equipment that is required to
respond to events while in Modes 4 and 5. There is no impact on the
plant safety analyses. This change is considered as an upgrade to
Technical Specifications that will improve consistency within the
Technical Specifications. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
(G) TSTF-306, Rev. 2, Traversing In-Core Probe Instrumentation
Specifications Requirements
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The addition of a note that the penetration flow path may be un-
isolated under administrative control simply provides
[[Page 57988]]
consistency with what is already allowed elsewhere in [the]
Technical Specifications. The isolation function of the TIP valves
are mitigative equipment. They do not create any increased
possibility of an accident since they are mitigative. Also, the
operation of the manual shear valves is unaffected by this activity.
The ability to manually isolate the TIP system by either the normal
isolation valve or the shear valve would be unaffected by the
inoperable instrumentation. Therefore, the same action as for manual
isolation Functions provides an appropriate level of 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.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
All systems, structures, and components previously required for the
mitigation of a transient remain capable of fulfilling their
intended design functions. The proposed changes have no adverse
effects on any safety-related system or component and do not
challenge the performance or integrity of any safety related system.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
The addition of a note that the penetration flow path may be un-
isolated under administrative control simply provides consistency
with what is already allowed elsewhere in Technical Specifications.
The ability to manually isolate the TIP system by either the normal
isolation valve or the shear valve would be unaffected by the
inoperable instrumentation. Therefore, the same action as for manual
isolation Functions provides an appropriate level of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
(H) TSTF-416, Rev. 0 Clarification of LPCI Operability during
Decay Heat Removal Operations
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change makes the Technical Specifications and their
Bases consistent in their consideration of an LPCI subsystem aligned
for decay heat removal being considered OPERABLE for ECCS. The LCO
3.5.1 and LCO 3.5.2 Bases state that a LPCI subsystem may be
considered OPERABLE during alignment and operation for decay heat
removal. As a result, no initiators to accidents previously
evaluated are affected and no mitigating equipment assumed in the
accidents previously evaluated are affected since the allowance for
LPCI being considered operable during these type of shutdown cooling
alignments or operations was the intent of the current technical
Specifications. Consequently, the probability or consequences of an
accident previous evaluated is not significantly increased.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
All systems, structures, and components previously required for the
mitigation of a transient remain capable of fulfilling their
intended design functions. The proposed changes have no adverse
effects on any safety-related system or component and do not
challenge the performance or integrity of any safety related system.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
The proposed change makes the Technical Specifications and their
Bases consistent in their consideration of an LPCI subsystem aligned
for decay heat removal being considered OPERABLE for ECCS. The LCO
3.5.1 and LCO 3.5.2 Bases state that an LPCI subsystem may be
considered OPERABLE during alignment and operation for decay heat
removal. As the operability requirements of the LPCI subsystem are
unaffected, the margin of safety is unaffected. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety.
(I) STF-17, Rev. 2, Containment Airlock Testing Frequency
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The containment airlock is considered as mitigative equipment.
Therefore, there are no impacts on the probability of accidents. The
proposed surveillance frequency assures that the interlock is
working such that there is no unintentional opening of both airlock
doors when containment is required. Because the interlock is assured
to be working, there will be no significant increase in the
consequences of an accident. There is no degradation in the ability
of the interlock to assure the containment integrity function is
maintained. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
All systems, structures, and components previously required for the
mitigation of a transient remain capable of fulfilling their
intended design functions. The proposed changes have no adverse
effects on any safety-related system or component and do not
challenge the performance or integrity of any safety related system.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
The frequency of 24 months for the interlock testing has been
demonstrated to be adequate with regards to the reliability of the
airlock. There is no impact on the leak testing requirements. There
is no affect on the plant safety analyses. Therefore, the proposed
change does not involve a significant reduction in a margin of
safety.
(J) TSTF-30, Rev. 3, TSTF-323, Rev. 0, TSTF-45, Rev. 2, TSTF-46,
Rev. 1, and TSTF-269, Rev. 2, Containment Isolation on Valve
Specification Changes
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The equipment affected by these changes is for mitigative
purposes. Therefore, there cannot be an increase in the probability
of occurrence of an accident. The controls required in the Technical
Specifications are adequate to ensure that the containment barriers
are ensured. Isolation valves will be assured to be in their correct
positions. Also, inoperable isolation valves in closed systems and
inoperable EFCVs have been evaluated to not have any significant
impact to the consequences of an accident due to the closed system
providing a barrier for the inoperable closed system isolation valve
and bounding analyses have been performed for EFCV instrument line
failures. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
All systems, structures, and components previously required for the
mitigation of a transient remain capable of fulfilling their
intended design functions. The proposed changes have no adverse
effects on any safety-related system or component and do not
challenge the performance or integrity of any safety related system.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
The equipment affected by these changes is for mitigative
purposes. The controls
[[Page 57989]]
required in the Technical Specifications are adequate to ensure that
the containment barriers are ensured. There is no affect on the
plant safety analyses. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
(K) STF-322, Rev. 2, Secondary Containment Operability
Clarification
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This change involves an administrative clarification to reflect
the original intent of the Technical Specifications. There is no
impact on the availability of the secondary containment.
Additionally, secondary containment is mitigative equipment.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
All systems, structures, and components previously required for the
mitigation of a transient remain capable of fulfilling their
intended design functions. The proposed changes have no adverse
effects on any safety-related system or component and do not
challenge the performance or integrity of any safety related system.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
This change involves an administrative clarification to reflect
the original intent of the Technical Specifications. There is no
impact on the availability of the secondary containment. There is no
impact on the plant safety analyses. Therefore, the proposed change
does not involve a significant reduction in a margin of safety.
(L) TSTF-276, Rev. 2, Power Factor for Emergency Diesel
Generator (EDG) Surveillences
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
These changes only affect mitigative equipment and therefore,
would not have an impact on the probability of an accident. Also,
the performance of the surveillances ensures that mitigative
equipment is capable of performing its intended function. Therefore,
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
All systems, structures, and components previously required for the
mitigation of a transient remain capable of fulfilling their
intended design functions. The proposed changes have no adverse
effects on any safety-related system or component and do not
challenge the performance or integrity of any safety related system.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
The performance of the surveillances ensures that mitigative
equipment is capable of performing its intended function. There are
no degradations in equipment readiness to mitigate design events.
There is no adverse affect on the plant safety analysis. Therefore,
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
(M) TSTF-65, Rev. 1, Generic Organizational Titles;
TSTF-299, Rev. 0, Primary Coolant Sources Inspection
Requirements;
TSTF-279, Rev. 0, Inservice Testing Program Clarifications;
TSTF-118, Rev. 0, and TSTF-106, Rev. 1, Diesel Generator Fuel
Oil Testing Program Clarifications;
TSTF-152, Rev. 0, Routine Reporting Requirement Upgrade
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The changes to Technical Specification 5.0, Administrative
Controls, are considered administrative changes. There are no
changes to plant structures, systems or components involved with
this change. There are no degradations in the availability of
mitigative plant equipment. The proposed changes provide
enhancements to the administrative controls in Technical
Specifications, therefore, there is no affect on any plant safety
analyses; therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, failure mechanisms, or limiting
single failures are introduced as a result of the proposed changes.
All systems, structures, and components previously required for the
mitigation of a transient remain capable of fulfilling their
intended design functions. The proposed changes have no adverse
effects on any safety-related system or component and do not
challenge the performance or integrity of any safety related system.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
(3) Does the proposed change involve a significant reduction in
a margin of safety?
Response: No.
The changes to Technical Specification 5.0, Administrative
Controls, are considered administrative changes. There are no
changes to plant structures, systems or components involved with
this change. There are no degradations in the availability of
mitigative plant equipment. The proposed changes provide
enhancements to the administrative controls in Technical
Specifications; therefore, there is no affect on any plant safety
analyses. 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 S. O'Neill, Associate and General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Daniel S. Collins, Acting.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: August 20, 2004.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) regarding the requirement to
demonstrate transfer of the unit A.C. electrical power supply to each
offsite circuit and would increase the surveillance exceptions for the
A.C. electrical sources in shutdown Modes 5 and 6. Also, the proposed
amendment would delete the TS requirement that the auto-connected loads
to each emergency diesel generator (EDG) do not exceed the 2000-hour
rating of the EDG.
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?
[[Page 57990]]
No. The proposed surveillance requirement changes do not alter
the design or operation of any structure, system, or component. No
previously analyzed accident scenario is changed. Initiating
conditions and assumptions remain as previously analyzed. The
revised surveillance requirements will continue to assure adequate
performance of structures, systems, and components. 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?
No. The proposed surveillance requirement changes do not alter
the design or operation of any structure, system, or component. No
new or different accident initiators are created as a result of the
proposed changes. 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?
No. The proposed surveillance requirement changes do not reduce
or adversely affect the capabilities of the offsite and onsite
electrical power sources. The revised surveillance requirements will
continue to assure adequate performance of structures, systems, and
components. The proposed changes do not affect conformance of the
electrical power systems to the applicable design criteria.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The Nuclear Regulatory Commission (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.
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 amendment request: September 1, 2004.
Description of amendment request: The proposed amendments would
revise the Operating Licenses' licensing basis to allow use of the code
for Generation of Thermal-Hydraulic Information for Containment,
Version 7.1patch1 (GOTHIC 7) to model Prairie Island Nuclear Generating
Plant (PINGP) containment response for loss of coolant accidents (LOCA)
and main steam line break (MSLB) accidents. The current PINGP
containment response analyses are performed utilizing CONTEMPT. The
Nuclear Management Company is making this request to support a
transition option from internal analyses using CONTEMPT to an external
analyses vendor (Westinghouse), which supports GOTHIC 7.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR), Part 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.
The proposed amendment will change the Prairie Island Nuclear
Generating Plant licensing basis by allowing use of the Generation
of Thermal-Hydraulic Information for Containment, Version 7.1patch1,
to model containment response for loss of coolant accident (LOCA)
and main steam line break (MSLB) accidents.
The containment is not an accident initiator, thus changing the
containment modeling methodology does not increase the probability
of an accident. This license amendment proposes to use a new
methodology for modeling containment response analyses following an
accident inside containment involving release of steam and water.
This amendment does not alter the nuclear reactor core or reactor
coolant system equipment, nor does it alter the methods or equipment
used directly in mitigation of an accident. Thus radioactive
releases inside containment due to an accident and radioactive
releases from containment are not affected by the proposed change in
analysis methodology. As discussed in Exhibits C and D, the Gothic 7
sample results for the LOCA and MSLB transients predicted that the
containment would remain below design pressure for both cases.
Therefore, this change does not increase the consequences of an
accident previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in 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 amendment will change the Prairie Island Nuclear
Generating Plant licensing basis by allowing use of the Generation
of Thermal-Hydraulic Information for Containment, Version 7.1patch1,
to model containment response for LOCA and MSLB accidents.
The proposed amendment does not involve changes to plant design,
hardware, system operation, or procedures involved with containment
function. The proposed changes include application of new
methodology for analysis of containment response following a loss of
coolant accident or steam line break accident. The results of the
analyses are used to demonstrate that the acceptance criteria for
the containment structure continue to be met. These changes do not
create the possibility for a new or different kind of accident.
Therefore, the proposed changes do 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 amendment will change the Prairie Island Nuclear
Generating Plant (PINGP) licensing basis by allowing use of the
Generation of Thermal-Hydraulic Information for Containment, Version
7.1patch1 (GOTHIC 7), to model containment response for LOCA and
MSLB accidents.
The proposed licensing basis change to use GOTHIC 7 affects the
design basis LOCA and MSLB containment accident analyses. As
discussed in Exhibits C and D, the GOTHIC 7 sample results for the
LOCA and MSLB transients predicted that the containment would remain
below design pressure for both cases. The GOTHIC 7 accuracy in this
application has been verified through benchmark analyses against the
current analyses of record, validated against recognized standard
data, and found to be appropriate for application to the PINGP
design basis accidents. Safety analysis acceptance criteria are
satisfied and adherence to safety analysis acceptance criteria using
GOTHIC 7 assures that Technical Specification limits will not be
exceeded during normal operation. 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 requests involve no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Section Chief: L. Raghavan.
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 amendment request: May 21, 2004.
Description of amendment request: The proposed amendment deletes
the requirements from the technical specifications (TS) to maintain
hydrogen recombiners and hydrogen monitors. Licensees were generally
required to implement upgrades as
[[Page 57991]]
described in NUREG-0737, ``Clarification of TMI [Three Mile Island]
Action Plan Requirements,'' and Regulatory Guide (RG) 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI Unit 2. Requirements related to
combustible gas control were imposed by Order for many facilities and
were added to or included in the TS for nuclear power reactors
currently licensed to operate. The revised 10 CFR 50.44, ``Standards
for Combustible Gas Control System in Light-Water-Cooled Power
Reactors,'' eliminated the requirements for hydrogen recombiners and
relaxed safety classifications and licensee commitments to certain
design and qualification criteria for hydrogen and oxygen monitors.
The NRC staff issued a notice of availability of a model no
significant hazards consideration 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 May 21, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen monitors are no longer required to mitigate design-basis
accidents and, therefore, the hydrogen monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
Category 1 in RG 1.97 is intended for key variables that most
directly indicate the accomplishment of a safety function for
design-basis accident events. The hydrogen monitors no longer meet
the definition of Category 1 in RG 1.97. As part of the rulemaking
to revise 10 CFR 50.44 the Commission found that Category 3, as
defined in RG 1.97, is an appropriate categorization for the
hydrogen monitors because the monitors are required to diagnose the
course of beyond design-basis accidents.
The regulatory requirements for the hydrogen monitors can be
relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3, and removal
of the hydrogen monitors from TS will not prevent an accident
management strategy through the use of the severe accident
management guidelines (SAMGs), the emergency plan (EP), the
emergency operating procedures (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: Thomas G. Eppink, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Section Chief (Acting): Mary Jane Ross-Lee.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: August 26, 2004.
Description of amendment requests: The proposed amendments would
revise the Technical Specifications (TS) to implement
ZIRLOTM fuel rod cladding material into the fuel design for
San Onofre Nuclear Generating Station (SONGS), Units 2 and 3.
Specifically, the licensee requests to add reference to
ZIRLOTM clad fuel and filler rods in TS 4.2.1, ``Fuel
Assemblies,'' and in TS 5.7.1.5, ``Core Operating Limits Report
(COLR),'' add the following references to the list of analytical
methods used to determine the core operating limits: ``Calculative
Methods for the C-E Nuclear Power Large Break LOCA [loss-of-coolant
accident] Evaluation Model,'' CENPD-1 32, Supplement 4-P-A, August
2000, and ``Implementation of ZIRLOTM Cladding Material in
CE [Combustion Engineering, Inc.] Nuclear Power Fuel Assembly
Designs,'' CENPD-404-P-A, November 2001.
[[Page 57992]]
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 allows the use of methods required for the
implementation of ZIRLOTM clad fuel rods in San Onofre
Nuclear Generating Station (SONGS) Units 2 and 3. The use of this
methodology will not increase the probability of an accident because
the plant systems will not be operated outside of design limits, no
different equipment will be operated, and system interfaces will not
change.
As ZIRLOTM material is introduced to the reactor,
transition cores will exist in which fuel assemblies containing
ZIRLOTM and Zircaloy clad fuel rods are co-resident. Each
type of fuel assembly (ZIRLOTM or Zircaloy clad fuel
rods) will be evaluated based on the approved topical reports listed
in TS 5.7.1.5.
The use of this additional methodology will not increase the
consequences of an accident because Limiting Conditions of Operation
(LCOs) will continue to restrict operation to within the regions
that provide acceptable results, and Reactor Protection System (RPS)
trip setpoints will restrict plant transients so that the
consequences of accidents will be acceptable. In addition, the
consequences of the accidents will be calculated using NRC accepted
methodologies.
The transition cores that will exist as ZIRLOTM clad
fuel is introduced to the reactor will not increase the consequences
of an accident. Operation within the LCOs and RPS setpoints will
continue to restrict plant transients so that the consequences of
accidents will be acceptable.
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 add any new equipment, modify any
interfaces with any existing equipment, alter the equipment's
function, or change the method of operating the equipment. The
proposed change does not alter plant conditions in a manner that
could affect other plant components. The proposed change does not
cause any existing equipment to become an accident initiator. The
ZIRLOTM clad fuel rod design does not introduce features
that could initiate an accident.
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.
Safety Limits ensure that Specified Acceptable Fuel Design
Limits (SAFDLs) are not exceeded during steady state operation,
normal operational transients and anticipated operational
occurrences. All fuel limits and design criteria shall be met based
on the approved methodologies defined in the topical reports. The
RPS in combination with the LCOs will continue to prevent any
anticipated combination of transient conditions for reactor coolant
system temperature, pressure, and thermal power level that would
result in a violation of the Safety Limits. Therefore, the proposed
changes will have no impact on the margins as defined in the
Technical Specification bases.
The safety analyses determine the LCO settings and RPS setpoints
that establish the initial conditions and trip setpoints, which
ensure that the Design Basis Events (Postulated Accidents and
Anticipated Operational Occurrences) analyzed in the Updated Final
Safety Analysis Report (UFSAR) produce acceptable results. In
addition, all fuel limits and design criteria shall be satisfied.
The Design Basis Events that are impacted by the implementation of
ZIRLOTM cladding will be analyzed using the NRC accepted
methodology described in CENPD-404-P-A.
The change in the fuel rod cladding material and the use of the
Emergency Core Cooling System (ECCS) performance evaluation models,
CENPD-132, Supplement 4-P-A, ``Calculative Methods for the CE
Nuclear Power Large Break LOCA Evaluation Model'' and CENPD-137,
Supplement 2-P-A, ``Calculative Methods for the ABB [Asea Brown
Boveri] CE Small Break LOCA Evaluation Model'' will not involve a
reduction in the margin of safety because LCOs and Limiting Safety
System Settings (LSSS) will be adjusted, if necessary, to maintain
acceptable results for the impacted Design Basis Events.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Robert Gramm.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: May 21, 2004.
Description of amendment request: The proposed amendment would
delete requirements from the Technical Specifications (TSs) to maintain
hydrogen recombiners (Unit 2 only) and hydrogen and oxygen monitors. A
notice of availability for this TS improvement using the consolidated
line item improvement process was published in the Federal Register on
September 25, 2003 (68 FR 55416). Licensees were generally required to
implement upgrades as described in NUREG-0737, ``Clarification of TMI
[Three Mile Island] Action Plan Requirements,'' and Regulatory Guide
1.97, ``Instrumentation for Light-Water-Cooled Nuclear Power Plants to
Assess Plant and Environs Conditions During and Following an
Accident.'' Implementation of these upgrades was an outcome of the
lessons learned from the accident that occurred at TMI, Unit 2.
Requirements related to combustible gas control were imposed by Order
for many facilities and were added to or included in the TSs for
nuclear power reactors currently licensed to operate. The revised 10
CFR 50.44, ``Standards for Combustible Gas Control System in Light-
Water-Cooled Power Reactors,'' eliminated the requirements for hydrogen
recombiners and relaxed safety classifications and licensee commitments
to certain design and qualification criteria for hydrogen and oxygen
monitors.
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on September
25, 2003 (68 FR 55416). The licensee affirmed the applicability of the
model NSHC determination in its application dated May 21, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not
[[Page 57993]]
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. Category 1 in RG 1.97 is intended for key variables that
most directly indicate the accomplishment of a safety function for
design-basis accident events. The hydrogen 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 TSs will not prevent an
accident management strategy through the use of the severe accident
management guidelines, the emergency plan, the emergency operating
procedures, and the site survey monitoring that support modification
of emergency plan protective action recommendations.
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen and oxygen monitor
requirements, including removal of these requirements from TSs does
not involve a significant increase in the probability or the
consequences of any accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From any Previously Evaluated
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen and oxygen monitor requirements,
including removal of these requirements from TSs 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 TSs, in light of
existing plant equipment, instrumentation, procedures, and programs
that provide effective mitigation of and recovery from reactor
accidents, results in a neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI, Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
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 TSs will not result in a
significant reduction in their functionality, reliability, and
availability.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: Mary Jane Ross-Lee, Acting.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: July 20, 2004.
Description of amendment request: The proposed amendments would
revise Administrative Controls Section 5.3.1 to replace the specific
designation for the Health Physics Superintendent with a reference to
the senior individual in charge of Health Physics, and to add
flexibility to the qualification requirements for unit staff positions.
This change supports Southern Nuclear Company's ongoing initiative to
achieve fleet standardization.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change to Technical Specifications Administrative
Controls Section 5.3.1 involves the use of a more generic
designation for the unit staff position responsible for Health
Physics without reducing the level of authority required for that
position. The proposed change also allows the flexibility to use an
NRC accredited program for qualifying personnel to fill unit staff
positions, which represents an acceptable alternative to the
qualification requirements for these positions as currently
specified in the Technical Specifications. Since the proposed
changes are administrative in nature, they do not involve any
physical changes to any structures, systems, or components, nor will
their performance requirements be altered. The proposed changes also
do not affect the operation, maintenance, or testing of the plant.
Therefore, the response of the plant to previously analyzed
accidents will not be affected. Consequently, 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 previously evaluated?
The proposed changes to the Technical Specifications will have
no adverse impact on the overall qualification of the unit staff.
The alternative use of an accredited program that has been endorsed
by the NRC will ensure the educational requirements and power plant
experience for each unit staff position are properly satisfied and
will continue to fulfill applicable regulatory requirements. Also,
since no change is being
[[Page 57994]]
made to the design, operation, maintenance, or testing of the plant,
no new methods of operation or failure modes are introduced by the
proposed changes. Therefore, the possibility of a new or different
kind of accident from any previously evaluated is not created.
(3) Does the proposed change involve a significant decrease in
the margin of safety?
The proposed changes to the Technical Specifications will have
no adverse impact on the onsite organizational features necessary to
assure safe operation of the plant. Lines of authority for plant
operation are unaffected by the proposed changes. Also, the adoption
of the more generic designation of the individual responsible for
Health Physics will reduce the regulatory burden of having to devote
limited resources to process a license amendment whenever a title
change for this position is implemented. Accordingly, this reduction
in regulatory burden and the option to use an accredited program
endorsed by NRC to qualify the unit staff will improve plant
efficiency without compromising plant safety. Therefore, the
proposed changes do not involve a significant decrease in the margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: Mary Jane Ross-Lee, Acting.
Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364,
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama
Date of amendment request: May 21, 2004.
Description of amendment request: The proposed amendment would
delete the requirements from the Technical Specifications (TS) to
maintain hydrogen recombiners and hydrogen monitors. A notice of
availability for this improvement using the consolidated line item
improvement process was published in the Federal Register on September
25, 2003 (68 FR 55416). Licensees were generally required to implement
upgrades as described in NUREG-0737, ``Clarification of TMI [Three Mile
Island] Action Plan Requirements,'' and Regulatory Guide (RG) 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI Unit 2. Requirements related to
combustible gas control were imposed by Order for many facilities and
were added to or included in the TSs for nuclear power reactors
currently licensed to operate. The revised 10 CFR 50.44, ``Standards
for Combustible Gas Control System in Light-Water-Cooled Power
Reactors,'' eliminated the requirements for hydrogen recombiners and
relaxed safety classifications and licensee commitments to certain
design and qualification criteria for hydrogen and oxygen monitors.
The NRC staff issued a notice of availability of a model no
significant hazards consideration determination (NSHC) 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 May 21, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen monitors are no longer required to mitigate design-basis
accidents and, therefore, the hydrogen monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
Category 1 in RG 1.97 is intended for key variables that most
directly indicate the accomplishment of a safety function for
design-basis accident events. The hydrogen monitors no longer meet
the definition of Category 1 in RG 1.97. As part of the rulemaking
to revise 10 CFR 50.44 the Commission found that Category 3, as
defined in RG 1.97, is an appropriate categorization for the
hydrogen monitors because the monitors are required to diagnose the
course of beyond design-basis accidents.
The regulatory requirements for the hydrogen monitors can be
relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3, and removal
of the hydrogen monitors from TSs will not prevent an accident
management strategy through the use of the severe accident
management guidelines, the emergency plan, the emergency operating
procedures, and site survey monitoring that support modification of
emergency plan protective action recommendations.
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen monitor requirements,
including removal of these requirements from 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 TSs, will not result in any failure mode
not previously analyzed. The hydrogen recombiner and hydrogen
monitor equipment was intended to mitigate a design-basis hydrogen
release. The hydrogen recombiner and hydrogen monitor equipment are
not considered accident precursors, nor does their existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within
the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from 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
[[Page 57995]]
approximately 24 hours after the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI, Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Therefore, this change does not involve a significant reduction
in the margin of safety. Removal of hydrogen monitoring from TS will
not result in a significant reduction in their functionality,
reliability, and availability.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Section Chief: Mary Jane Ross-Lee, Acting.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of amendment request: May 21, 2004.
Description of amendment request: The proposed amendment would
delete the requirements from the Technical Specifications (TSs) to
maintain hydrogen recombiners and hydrogen monitors. A notice of
availability for the TS improvement using the consolidated line item
improvement process was published in the Federal Register on September
25, 2003 (68 FR 55416). Licensees were generally required to implement
upgrades as described in NUREG-0737, ``Clarification of TMI [Three Mile
Island] Action Plan Requirements,'' and Regulatory Guide (RG) 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI, Unit 2. Requirements related to
combustible gas control were imposed by Order for many facilities and
were added to or included in the TSs for nuclear power reactors
currently licensed to operate. The revised 10 CFR 50.44, ``Standards
for Combustible Gas Control System in Light-Water-Cooled Power
Reactors,'' eliminated the requirements for hydrogen recombiners and
relaxed safety classifications and licensee commitments to certain
design and qualification criteria for hydrogen and oxygen monitors.
The NRC staff issued a notice of availability of a model no
significant hazards consideration determination (NSHC) 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 May 21, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen monitors are no longer required to mitigate design-basis
accidents and, therefore, the hydrogen monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
Category 1 in RG 1.97 is intended for key variables that most
directly indicate the accomplishment of a safety function for
design-basis accident events. The hydrogen monitors no longer meet
the definition of Category 1 in RG 1.97. As part of the rulemaking
to revise 10 CFR 50.44, the Commission found that Category 3, as
defined in RG 1.97, is an appropriate categorization for the
hydrogen monitors because the monitors are required to diagnose the
course of beyond design-basis accidents.
The regulatory requirements for the hydrogen monitors can be
relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3, and removal
of the hydrogen monitors from TSs will not prevent an accident
management strategy through the use of the severe accident
management guidelines, the emergency plan, the emergency operating
procedures, and the site survey monitoring that support modification
of emergency plan protective action recommendations.
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen monitor requirements,
including removal of these requirements from TSs, does not involve a
significant increase in the probability or the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TSs, will not result in any failure mode
not previously analyzed. The hydrogen recombiner and hydrogen
monitor equipment was intended to mitigate a design-basis hydrogen
release. The hydrogen recombiner and hydrogen monitor equipment are
not considered accident precursors, nor does their existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within
the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TSs, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI, Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Therefore, this change does not involve a significant reduction
in the margin of safety. Removal of hydrogen monitoring from TSs
will not result in a significant reduction in
[[Page 57996]]
their functionality, reliability, and availability.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Section Chief: Mary Jane Ross-Lee, Acting.
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 26, 2004.
Description of amendment request: The license amendment request
proposes revising the Technical Specifications (TSs) to delete the TS
requirements related to Hydrogen Analyzers and Hydrogen Recombiners
consistent with NRC-approved TS Task Force (TSTF) Traveler number TSTF-
447, Revision 1, ``Elimination of Hydrogen Recombiners and Change to
Hydrogen and Oxygen Monitors.'' The TS requirements related to Hydrogen
Analyzers and Hydrogen Recombiners are contained in TS Tables 3.3-10
and 4.3-10 and TSs 3.6.4.1 and 3.6.4.2. The availability of this TS
improvement was announced in the Federal Register on September 25,
2003, as part of the Consolidated Line Item Improvement Process
(CLIIP).
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 analysis endorses the NRC staff's generic no
significant hazards consideration determination for TSTF-447 which was
published in the Federal Register on September 25, 2003 (68 FR 55416)
as follows:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design basis LOCA hydrogen release,
hydrogen [and oxygen] monitors are no longer required to mitigate
design-basis accidents and, therefore, the hydrogen monitors do not
meet the definition of a safety-related component as defined in 10
CFR 50.2. RG [Regulatory Guide] 1.97 Category 1, is intended for key
variables that most directly indicate the accomplishment of a safety
function for design-basis accident events. The hydrogen 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.
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.
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
[[Page 57997]]
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 e-mail to [email protected].
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendment: August 6, 2002, as supplemented
December 12, 2002, July 24, 2003, and March 1, May 20, and August 11,
2004.
Brief description of amendment: The amendments replace the
Technical Specifications 3.9.4 and 3.9.5 requirements to close all
containment penetrations providing direct access from the containment
atmosphere to outside temperature with a set of more detailed and less
restrictive requirements.
Date of issuance: September 13, 2004.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 268 and 244.
Renewed Facility Operating License No. DPR-53: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: October 15, 2002 (67 FR
63690).
The December 12, 2002, July 24, 2003, March 1, 2004, and May 20,
2004, letters provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination. The August 11, 2004,
letter withdrew the licensee's requested changes to Technical
Specification 3.9.3.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 13, 2004.
No significant hazards consideration comments received: No.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: December 12, 2003.
Brief description of amendments: The amendments delete Technical
Specification Section 5.5.3, ``Post-Accident Sampling.''
Date of issuance: September 15, 2004.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 269 and 245.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19564).
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated September 15, 2004.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: May 7, 2002, as supplemented
April 7, 2003 and July 19, 2004.
Brief description of amendment: The amendment relocates the
boration system Technical Specification (TS) requirements to the
Technical Requirements Manual and the boron dilution analysis
restrictions within the TSs. The amendment also revises the TS limiting
condition for operation action and the surveillance requirements
associated with the emergency core cooling, containment spray and
cooling and auxiliary feedwater systems.
Date of issuance: September 9, 2004.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 283.
Facility Operating License No. DRP-65: The amendment revised the
TSs.
Date of initial notice in Federal Register: June 11, 2002 (67 FR
40021). The April 7, 2003, and July 19, 2004, supplements contained
clarifying information and did not change the staff's initial proposed
finding of no significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 9, 2004.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: August 7, 2002, as supplemented
November 5, 2003.
Brief description of amendment: The amendment revises the Technical
Specifications (TSs) related to safety system settings. Specifically,
the amendment revises: (1) TS 1.0 ``Definitions;'' (2) TS 2.2.1
``Limiting Safety System Settings--Reactor Trip System Instrumentation
Setpoints;'' (3) TS 3.3.1 ``Reactor Trip System Instrumentation;'' (4)
TS 3.3.2 ``Engineered Safety Features Actuation System
Instrumentation;'' (5) TS 3.7.7 ``Control Room Emergency Ventilation
System;'' and (6) TS 3.8.3.1 ``Onsite Power Distribution--Operating.''
Date of issuance: September 14, 2004.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 220.
Facility Operating License No. DRP-49: The amendment revised the
TSs.
Date of initial notice in Federal Register: October 15, 2002 (67 FR
63692). The November 5, 2003, supplement contained clarifying
information and did not change the staff's initial proposed finding of
no significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 14, 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: May 25, 2004.
[[Page 57998]]
Brief description of amendments: The amendments revised the
licensing basis in the Updated Final Safety Analysis Report (UFSAR) to
support installation of a low-pressure injection (LPI) cross connect
inside containment. The changes to the UFSAR revise the licensing basis
for selected portions of the core flood and LPI/Decay Heat Removal
piping to allow exclusion of the dynamic effects associated with
postulated rupture of that piping by application of leak-before-break
technology. The amendments also revise the Technical Specifications
(TSs) to delete TSs that will no longer apply when the LPI cross
connect modification has been implemented.
Date of issuance: September 2, 2004.
Effective date: As of the date of issuance and shall be implemented
during the fall 2004 refueling outage of Unit 3.
Amendment Nos.: 340, 342, and 341.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the TSs.
Date of initial notice in Federal Register: July 6, 2004 (69 FR
40673). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 2, 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: January 20, 2004, as
supplemented by letters dated May 19, July 13, and August 16, 2004.
Brief description of amendments: The amendments change the Prairie
Island technical specification (TS) on containment to implement a
portion of TSs Task Force Traveler 5, ``Revise containment requirements
during handling irradiated fuel and core alterations.'' The amendments
also selectively implement an alternative source term per Title 10 of
the Code of Federal Regulations, Section 50.67 to perform the
radiological consequences analysis of the design-basis fuel handling
accident which supports the proposed TS changes.
Date of issuance: September 10, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 166 and 156.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 25, 2004 (69 FR
29769 ).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 10, 2004.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California
Date of application for amendment: June 23, 2004.
Brief description of amendment: The amendment removes a restriction
from the Humboldt Bay Power Plant Unit 3 license thereby permitting
Pacific Gas and Electric to engage in active decommissioning of the
facility.
Date of issuance: September 10, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 35.
Facility Operating License No. DPR-7: This amendment revises the
license.
Date of initial notice in Federal Register: August 3, 2004 (69 FR
46587).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 10, 2004.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: June 28, 2004, as supplemented by
letter dated August 5, 2004.
Brief Description of amendments: The amendments revise TS 3.4.13,
``RCS [Reactor Coolant System] Operational Leakage,'' TS 5.5.9, ``Steam
Generator [SG] Tube Surveillance Program,'' and TS 5.6.10, ``Steam
Generator Tube Inspector Report.'' They also add a new TS 3.4.17,
``Steam Generator Tube Integrity.'' These changes facilitate
implementation of industry initiative NEI [Nuclear Energy Institute]
97-08, ``Steam Generator Program Guidelines,'' which allows a
comprehensive, performance-based approach to managing SG performance at
Farley Nuclear Plant, Units 1 and 2.
Date of issuance: September 10, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 163 and 156.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: August 3, 2004 (69 FR
46950). The supplemental letter dated August 5, 2004, provided
clarifying information that did not change the initial proposed no
significant hazards consideration determinations.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 10, 2004.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: September 6, 2002, as
supplemented by letters dated December 19, 2002, March 28, June 24,
September 3, and October 22, 2003.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to (1) relocate the pressure temperature limit
curves and low temperature overpressure protection system limits to the
Pressure and Temperature Limits Report (PTLR), (2) reference the PTLR
in the affected TSs limiting conditions for operation and bases,
including the addition of the PTLR to the definitions section of the
TSs, and the addition of a new TS 6.9.1.15 to the administrative
controls section of the TSs, (3) relocate TS 3.4.9.2, Pressurizer, to
the Sequoyah Technical Requirements Manual and (4) revise TS 3.4.9.1,
Pressure/Temperature Limits, Reactor Coolant System, and TS 3.4.12, Low
Temperature Over Pressure Protection Systems, to incorporate standard
TSs requirements from NUREG-1431, Revision 2, ``Standard Technical
Specifications--Westinghouse Plants.''
Date of issuance: September 15, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 294 and 284.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 29, 2002 (67 FR
66015). The supplemental letters provided clarifying information that
did not expand the scope of the original application or change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a
[[Page 57999]]
Safety Evaluation dated September 15, 2004.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 17th day of September, 2004.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 04-21345 Filed 9-27-04; 8:45 am]
BILLING CODE 7590-01-P