[Federal Register Volume 69, Number 81 (Tuesday, April 27, 2004)]
[Notices]
[Pages 22877-22890]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-9225]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
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 April 2, 2004, through April 15, 2004. The
last biweekly notice was published on April 13, 2004 (69 FR 19561).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor
[[Page 22878]]
intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) e-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to 301-415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the 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].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: March 19, 2004.
Description of amendment request: The licensee proposed to revise
Section 4.2, ``Reactivity Control,'' of the Technical Specifications.
Specifically, the amendment would revise Subsection 4.2.C, regarding
surveillance requirements associated with control rod scram time
testing (STT) by: (1) Eliminating unnecessary depressurized STT of non-
maintenance-affected control rods, (2) providing the required STT data
necessary to apply actual scram times to implement improved minimum
critical power ratio operating limits, and (3) eliminating the
resulting redundant requirement to test ``eight control rods'' after a
reactor scram or other outage. The amendment will also include
editorial and pagination changes to accommodate the proposed technical
changes.
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 adds new surveillance requirements (SR) to
the Minimum Critical Power Ratio (MCPR) Technical Specification (TS)
which requires determination of the MCPR operating limit following
the completion of scram time testing (STT) of the control rods. Use
of the scram speed in determining the MCPR operating limit (i.e.,
Option B) is an alternative to the current method for determining
the operating limit (i.e., Option A). The probability of an accident
previously evaluated is unrelated to the MCPR operating limit that
is provided to ensure no fuel damage results during anticipated
operational occurrences. This is an operational limit to ensure
conditions following an assumed accident do not result in fuel
failure and therefore do not contribute to the occurrence of an
accident. The proposed change eliminates unnecessary depressurized
STT of non-maintenance[-]affected control rods and the requirement
to test ``eight selected rods'' after a reactor scram or other
outage. The requirement to test ``eight selected rods'' is replaced
by a new SR to perform periodic STT. No active or passive failure
mechanisms that could lead to an accident are affected by this
proposed change. Therefore, the proposed change in STT requirements
does not significantly increase the probability of an accident
previously evaluated.
The proposed change ensures that the appropriate MCPR operating
limit is in place. By implementing the correct MCPR operating limit
the MCPR safety limit will continue to be ensured. Ensuring the MCPR
safety limit is not exceeded will result in prevention of fuel
failure. Therefore, since there is no increase in the potential for
fuel failure there is no increase in the consequences of any
accidents previously evaluated.
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 adds a new SR to the MCPR TS which requires
determination of the MCPR operating limit following the completion
of scram time testing of the control rods. The proposed change
eliminates unnecessary depressurized STT of non-maintenance[-
]affected rods and the requirement to test ``eight selected rods''
after a reactor scram or other outage. The
[[Page 22879]]
requirement to test ``eight selected rods'' is replaced by a new SR
to perform periodic STT. The proposed change does not involve the
use or installation of new equipment. Installed equipment is not
operated in a new or different manner. No new or different system
interactions are created, and no new processes are introduced. No
new failures have been created by the addition of the proposed SR
and the use of the alternate method for determining the MCPR
operating limit.
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.
Use of Option B for determining the MCPR operating limit will
result in a reduced operating limit in comparison to the use of
Option A. However, a reduction in the operating limit margin does
not result in a reduction in the safety margin. The MCPR safety
limit remains the same regardless of the method used for determining
the operating limit. The proposed change eliminates unnecessary
depressurized STT of non-maintenance[-]affected control rods and the
requirement to test ``eight selected rods'' after a reactor scram or
other outage. The requirement to test ``eight selected rods'' is
replaced by a new SR to perform periodic STT. No active or passive
failure mechanisms that could adversely impact the consequences of
an accident are affected by this proposed change. All analyzed
transient results remain well within the design values for
structures, systems and components.
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 General
Counsel, Exelon Generation Company, LCC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Richard J. Laufer.
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: March 23, 2004.
Brief description of amendments: The licensee proposed to revise
the Technical Specifications (TSs) by eliminating the requirements for
hydrogen/oxygen monitors. The proposed amendment supports
implementation of the revision to 10 CFR 50.44, ``Standards for
Combustible Gas Control System in Light-Water-Cooled Power Reactors,''
that became effective on October 16, 2003.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-447, ``Elimination of Hydrogen
Recombiners and Change to Hydrogen and Oxygen Monitors.'' The
availability of this TS improvement was published in the Federal
Register on September 25, 2003 (68 FR 55416), on possible amendments
concerning TSTF-447, including a model safety evaluation and model no
significant hazards consideration (NSHC) determination, using the
consolidated line item improvement process. In its application for
amendment, the licensee affirmed the applicability of the following
NSHC determination.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee presented
an analysis of NSHC by endorsing the model NSHC determination published
in 68 FR 55416 (reproduced below):
Criterion 1.--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen and oxygen monitors are no longer required to mitigate
design-basis accidents and, therefore, the hydrogen monitors do not
meet the definition of a safety-related component as defined in 10
CFR 50.2. RG [Regulatory Guide] 1.97 Category 1, is intended for key
variables that most directly indicate the accomplishment of a safety
function for design-basis accident events. The hydrogen and oxygen
monitors no longer meet the definition of Category 1 in RG 1.97. As
part of the rulemaking to revise 10 CFR 50.44 the Commission found
that Category 3, as defined in RG 1.97, is an appropriate
categorization for the hydrogen monitors because the monitors are
required to diagnose the course of beyond design-basis accidents.
Also, as part of the rulemaking to revise 10 CFR 50.44, the
Commission found that Category 2, as defined in RG 1.97, is an
appropriate categorization for the oxygen monitors, because the
monitors are required to verify the status of the inert containment.
The regulatory requirements for the hydrogen and oxygen monitors
can be relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3,
classification of the oxygen monitors as Category 2 and removal of
the hydrogen and oxygen monitors from TS will not prevent an
accident management strategy through the use of the 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 and oxygen monitor
requirements, including removal of these requirements from TS, does
not involve a significant increase in the probability or the
consequences of any accident previously evaluated.
Criterion 2.--The Proposed Change Does Not Create the
Possibility of a New or Different Kind of Accident From Any
Previously Evaluated.
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen and oxygen monitor requirements,
including removal of these requirements from TS, will not result in
any failure mode not previously analyzed. The hydrogen recombiner
and hydrogen and oxygen monitor equipment was intended to mitigate a
design-basis hydrogen release. The hydrogen recombiner and hydrogen
and oxygen monitor equipment are not considered accident precursors,
nor does their existence or elimination have any adverse impact on
the pre-accident state of the reactor core or post[-]accident
confinement of radionuclides within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3.--The Proposed Change Does Not Involve a Significant
Reduction in [a] Margin of Safety.
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen and oxygen monitor requirements,
including removal of these requirements from TS, in light of
existing plant equipment, instrumentation, procedures, and programs
that provide effective mitigation of and recovery from reactor
accidents, results in a neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-
[[Page 22880]]
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 [Three Mile Island],
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 [a] margin of safety. The intent of the requirements established
as a result of the TMI, Unit 2 accident can be adequately met
without reliance on safety-related oxygen monitors. Removal of
hydrogen and oxygen monitoring from TS will not result in a
significant reduction in their functionality, reliability, and
availability.
The NRC staff 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 General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Richard J. Laufer.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: November 12, 2002, as supplemented
March 5, 2004. This notice supersedes the notice that was published on
February 18, 2003 (68 FR 7813).
Description of amendments request: The proposed amendments would
revise the Technical Specifications to support an expansion of the core
flow operating range, including the new automated backup stability
protection 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:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change will implement DSS-CD [Detect and Suppress
Solution--Confirmation Density] as the long-term stability solution.
The DSS-CD solution is designed to identify the power oscillation
upon inception and initiate control rod insertion to terminate the
oscillations prior to any significant amplitude growth. The DSS-CD
provides protection against violation of the Safety Limit Minimum
Critical Power Ratio (SLMCPR) for anticipated oscillations.
Compliance with General Design Criteria (GDC) 10 and 12 of 10 CFR
part 50, Appendix A is accomplished via an automatic action. The
DSS-CD introduces an enhanced detection algorithm that detects the
inception of power oscillations and generates an earlier power
suppression trip signal exclusively based on successive period
confirmation recognition. The existing Option III algorithms are
retained, with generic setpoints, to provide defense-in-depth
protection for unanticipated reactor instability events.
A developing instability event is suppressed by the DSS-CD
system with substantial margin to the SLMCPR and no clad damage,
with the event terminating in a scram and never developing into an
accident. In addition, the DSS-CD solution defense-in-depth features
incorporate all the backup scram algorithms plus the licensed scram
feature of the existing Option III system. The DSS-CD system does
not interact with equipment whose failure could cause an accident.
Scram setpoints in the DSS-CD will be established so that analytical
limits are met. The reliability of the DSS-CD will meet or exceed
that of the existing system. No new challenges to safety-related
equipment will result from the DSS-CD solution. Because an
instability event would reliably terminate in an early scram without
impact on other safety systems, there is no significant increase in
the probability of an accident.
The existing requirement to initiate an alternate (i.e., manual)
method to detect and suppress thermal hydraulic instability
oscillations is expanded to include a requirement to either
implement an Automated Backup Stability Protection (ABSP) (i.e.,
Required Action I.2.1) or exit the operating region most susceptible
to rapid onset of Thermal Hydraulic Instability (THI) (i.e.,
Required Action I.2.2). The ABSP is an automatic reactor scram
region, implemented by the Average Power Range Monitor (APRM) flow-
biased scram setpoint. It may be used if the Oscillation Power Range
Monitoring (OPRM) system is inoperable to allow continued operation
within the MELLLA+ [Maximum Extended Load Line Limit Analysis Plus]
operating domain. Additionally, a new Required Action I.3 is
included. Required Action I.3 ensures that a report is made to the
NRC, if DSS-CD is inoperable for 120 days.
To maintain the existing margin between equipment operability
requirements and the region of power-flow operation where
anticipated events could lead to thermal-hydraulic instability, (1)
TS 3.3.1.1, Required Action J.1 is revised to require the plant to
be < 18% RTP [rated thermal power] versus < 20% RTP in the event
that the OPRM Upscale Function is inoperable and the Required
Actions associated with Action I are not completed, and (2) the
operability requirement for the OPRM Upscale Function (i.e., TS
3.3.1.1, Table Function 2.f) is changed from = 20% RTP to
= 18% RTP. This 5% margin is consistent with and
maintains the existing 5% margin operability requirements for the
Option III OPRM Upscale operability requirements.
Overall, these changes result in more conservative plant
operation. Other changes proposed in this supplement are either in
direct support of ABSP or are administrative in nature.
Proper operation of the DSS-CD system does not affect any
fission product barrier or Engineered Safety Feature. Thus, the
proposed change cannot change the consequences of any accident
previously evaluated. As stated above, the DSS-CD solution meets the
requirements of GDC 10 and 12 by automatically detecting and
suppressing design basis thermal-hydraulic oscillations prior to
exceeding the fuel SLMCPR.
Based on the above, the operation of the DSS-CD solution within
the framework of the Option III OPRM hardware will not increase 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 DSS-CD solution operates within the existing Option III OPRM
hardware. No new operating mode, safety-related equipment lineup,
accident scenario, system interaction, or equipment failure mode was
identified. The ABSP automatic reactor scram region is implemented
by adjusting the existing APRM flow-biased scram setpoint.
Therefore, the DSS-CD solution will not adversely affect plant
equipment.
Because there are no hardware changes, there is no change in the
possibility or consequences of a failure. The worst case failure of
the equipment is a failure to initiate mitigating action (i.e.,
scram), but no failure can cause an accident of a new or different
kind than any previously evaluated.
Based on the above, the proposed change to the DSS-CD solution
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The DSS-CD solution is designed to identify the power
oscillation upon inception and initiate control rod insertion to
terminate the oscillations prior to any significant amplitude
growth. The DSS-CD solution algorithm will maintain or increase the
margin to the SLMCPR for anticipated instability events. The safety
analyses in ``Detect And Suppress Solution--Confirmation Density
Licensing Topical Report,'' Revision 3 demonstrate the margin to the
SLMCPR for postulated bounding stability events. Existing margin
between equipment operability requirements and the region of power-
flow operation where
[[Page 22881]]
anticipated events could lead to thermal-hydraulic instability are
maintained. As a result, there is no impact on the SLMCPR identified
for an instability event.
The existing requirement to initiate an alternate method to
detect and suppress thermal hydraulic instability oscillations is
expanded to include a requirement to either implement an ABSP (i.e.,
Required Action I.2.1) or exit the operating region most susceptible
to rapid onset of THI (i.e., Required Action I.2.2). Additionally, a
new Required Action I.3 is included. Required Action I.3 ensures
that a report is made to the NRC, if DSS-CD is inoperable for 120
days. These change results in more conservative plant operation.
Other changes proposed in this supplement are either in direct
support of ABSP or are administrative in nature.
The current Option III algorithms (i.e., Period Based Detection,
Amplitude Based, and Growth Rate) are retained (with generic
setpoints) to provide defense-in-depth protection for unanticipated
reactor instability events.
Based on the above, the proposed change will 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: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: William F. Burton, Acting.
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: March 25, 2004.
Description of amendment request: The proposed amendments would
allow the use of the methodology described in Framatome-ANP (FRA-ANP)
Topical BAW-10169-A ``RSG Plant Safety Analysis--B&W Safety Analysis
Methodology for Recirculating Steam Generator Plants'', dated October
1989 for the generation of mass and energy release rates during a Main
Steam Line Break accident for Prairie Island Nuclear Generating Plant.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. 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 the use of the
methodology described in Framatome-ANP Topical BAW-10169-A ``RSG
Plant Safety Analysis--B&W Safety Analysis Methodology for
Recirculating Steam Generator Plants'' that utilizes the RELAP5/
MOD2-B&W code described in Topical BAW-10164-A ``RELAP5/MOD2-B&W--An
Advanced Computer Program for Light-Water Reactor LOCA [loss-of-
coolant accident] and Non-LOCA Transient Analysis'' for the
generation of predicted mass and energy releases during a Main Steam
Line Break accident.
The methodology used to perform an analysis of a main steam line
break is not an accident initiator, thus changing the methodology
does not increase the probability of an accident.
The mass and energy releases generated by the proposed
methodology will be utilized to demonstrate that the design basis
limits for fission product barriers are not exceeded. The proposed
methodology does not alter the nuclear reactor core, reactor coolant
system, or equipment used directly in mitigation of a main steam
line break, thus radioactive releases due to a main steam line break
accident are not affected by the proposed change in analysis
methodology. 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 the use of the
methodology described in Framatome-ANP Topical BAW-10169-A ``RSG
Plant Safety Analysis--B&W Safety Analysis Methodology for
Recirculating Steam Generator Plants'' that utilizes the RELAP5/
MOD2-B&W code described in Topical BAW-10164-A ``RELAP5/MOD2-B&W--An
Advanced Computer Program for Light-Water Reactor LOCA and Non-LOCA
Transient Analysis'' for the generation of predicted mass and energy
releases during a Main Steam Line Break accident.
The analysis of a main steam line break using the proposed
methodology does not alter the nuclear reactor core, reactor coolant
system, or equipment used directly in mitigation of a main steam
line break.
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 licensing basis by allowing the use of the
methodology described in Framatome-ANP Topical BAW-10169-A ``RSG
Plant Safety Analysis--B&W Safety Analysis Methodology for
Recirculating Steam Generator Plants'' that utilizes the RELAP5/
MOD2-B&W code described in Topical BAW-10164-A ``RELAP5/MOD2-B&W--An
Advanced Computer Program for Light-Water Reactor LOCA and Non-LOCA
Transient Analysis'' for the generation of predicted mass and energy
releases during a Main Steam Line Break accident.
The proposed licensing basis change will result in a
conservative calculation of the mass and energy releases during a
Main Steam Line Break accident. This will ensure that there is no
reduction in the margin of safety for analyses that utilize the
generated mass and energy releases as inputs. 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.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: March 4, 2004.
Description of amendment request: The proposed amendment would
revise the SSES 1 and 2 Technical Specification Table 3.3.5.1-1 to
clarify that four low pressure coolant injection pump discharge
pressure-high channels are required for each automatic depressurization
system trip 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:
1. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
Response: No.
The Technical Specification required number of protection
channels is not an initiator to any accident sequence analyzed in
the Final Safety Analysis Report (FSAR). As discussed in this
request, the change is editorial and involves no change in the
number of ADS [Automatic Depressurization System] supporting
protection channels
[[Page 22882]]
required by the Susquehanna Steam Electric Station (SSES) Technical
Specifications (TS). The change does not have any effect on the
initiator of any accident sequence analyzed in the Final Safety
Analysis Report (FSAR) and does not affect any assumptions
associated with the mitigation of accident or transient events. The
change does not involve any physical change to structures, systems,
or components (SSCs) and does not involve any physical change to
structures, systems, or components (SSCs) and does not alter the
method of operation or control of SSCs. The current assumptions in
the SSES FSAR safety analysis regarding accident initiators and
mitigation of accidents are unaffected by these changes. No
additional failure modes or mechanisms are being introduced and the
likelihood of previously analyzed failures remains unchanged.
Operation in accordance with the proposed Technical
Specification (TS) continues to ensure that the plant response to
analyzed accidents remains capable of performing as described in the
FSAR. Therefore, the mitigative functions supported by the system
continue to provide the protection assumed by the analysis.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. There
are no setpoints, at which protective or mitigative actions are
initiated, affected by this change. This change does not alter the
manner in which equipment operation is initiated, nor are the
function demands on credited equipment be[ing] changed. No
alterations in the procedures that ensure the plant remains within
analyzed limits are being proposed, and no changes are being made to
the procedures relied upon to respond to an off-normal event as
described in the FSAR. As such, no new failure modes are being
introduced. The change does not alter the assumptions made in the
safety analysis and licensing basis.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The change is editorial and involves no technical
changes to the Susquehanna Steam Electric Station (SSES) Technical
Specifications (TS). Therefore the plant response to analyzed events
continues to provide the margin of safety assumed by the analysis.
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: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: Richard J. Laufer.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: March 5, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification Surveillance Requirement (SR) 3.6.4.1.3
to require that only one secondary containment access door in each
access opening be verified closed. In addition, this SR allows entry
and exit access between required secondary containment zones that have
a single door.
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 of occurrence or consequences of an accident
previously evaluated?
Response: No.
The Technical Specification Surveillance being revised, which
verifies the status of the secondary containment access doors, is
not an initiator to any accident sequence analyzed in the Final
Safety Analysis Report (FSAR). The proposed change relaxes the
acceptance criteria of this Surveillance such that maintenance on
one of two airlock access doors can be performed. However, requiring
that at least one door is closed, in conjunction with the continued
requirement to maintain the building at a negative pressure,
continues to assure that the secondary containment barrier is
maintained operable. This provides adequate assurance that the
secondary containment is capable of performing the accident
mitigation function assumed in the accident analyses. As a result,
the consequences of any accident previously evaluated are not
significantly affected.
The Note, which was added to the Technical Specifications,
provides clarification and precludes a conflict with the explicit
wording of SR 3.6.4.1.3. Since this Note is consistent with the
intent as reflected in the Bases and with the prior SSES Technical
Specifications, the change is considered editorial and reflects an
administrative presentation preference and not a technical change.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant. No new equipment is being introduced, and installed
equipment is not being operated in a new or different manner. There
are no setpoints, at which protective or mitigative actions are
initiated, affected by this change. This change does not alter the
manner in which equipment operation is initiated, nor are the
function demands on credited equipment changed. No alterations in
the procedures that ensure the plant remains within analyzed limits
are being proposed, and no changes are being made to the procedures
relied upon to respond to an off-normal event as described in the
FSAR. As such, no new failure modes are being introduced.
The Note, which was added to the Technical Specifications,
provides clarification and precludes a conflict with the explicit
wording of SR 3.6.4.1.3. Since this Note is consistent with the
intent as reflected in the Bases and with the prior SSES Technical
Specifications, the change is considered editorial and reflects an
administrative presentation preference and not a technical change.
The change does not alter the assumptions made in the safety
analysis and licensing basis.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The margin of safety is established through equipment design,
operating parameters, and the setpoints at which automatic actions
are initiated. The change could allow additional time for one of two
airlock doors to be open for maintenance. However, the margin of
safety is maintained by the continued closure of the remaining
airlock door (as is currently allowed for normal entry and exit) and
the continued requirement to be able to maintain the building at a
negative pressure.
The Note, which was added to the Technical Specifications,
provides clarification and precludes a conflict with the explicit
wording of SR 3.6.4.1.3. Since this Note is consistent with the
intent as reflected in the Bases and with the prior SSES Technical
Specifications, the change is considered editorial and reflects an
administrative presentation preference and not a technical change.
Therefore, the plant response to analyzed events continues to
provide the margin of safety assumed by the analysis.
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: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: Richard J. Laufer.
[[Page 22883]]
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant (BFN), Units 2 and 3, Limestone County,
Alabama
Date of amendment request: July 31, 2002, as supplemented by
letters dated December 9, 2002, February 12, 2003, March 26, 2003, July
11, 2003, and July 17, 2003.
Description of amendment request: The proposed amendments request
full implementation of an alternative source term (AST) for the Units
1, 2, and 3 operating licenses. The amendments adopt the AST
methodology by revising the current accident source term and replacing
it with an accident source term as prescribed in 10 CFR 50.67. The
submittals also propose to revise/delete the Technical Specification
(TS) Sections associated with control emergency ventilation (CREV),
standby gas treatment (SGT), standby liquid control (SLC), and
secondary containment systems. Additionally, the submittals request
modification of the licensing and design basis to reflect the
application of the AST methodology and the function of the SLC system,
and deletion of a license condition for Units 2 and 3, which all the
actions have been completed.
The supplements to the original application include the withdrawal
of the request to delete one of the TS Sections described above,
associated with the absorption of elemental iodine by the SGT and CREV
systems charcoal filters. Also the supplements add a new TS Section to
require verification that the minimum fuel decay period has passed
prior to moving fuel after the reactor is shut down. The licensee
indicated that these modifications/deletions do not affect the
originally published no significant hazards consideration. The original
no hazards consideration is reproduced below.
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. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The AST and those plant systems affected by implementing AST do
not initiate DBAs [design-basis accidents]. The AST does not affect
the design or operation of the facility; rather, once the occurrence
of an accident has been postulated, the new source term is an input
to evaluate the consequences. The implementation of the AST has been
evaluated in the analyses for the limiting DBAs at BFN. The
equipment affected by the proposed change is mitigative in nature
and relied upon following an accident. The proposed changes to the
TS do revise certain performance requirements. However, these
changes will not involve a revision to the parameters or conditions
that could contribute to the initiation of a design basis accident
discussed in Chapter 14 of the BFN Updated Final Safety Analysis
Report.
Plant specific radiological analyses have been performed and,
based on the results of these analyses, it has been demonstrated
that the dose consequences of the limiting events considered in the
analyses are within the regulatory guidance provided by the NRC for
use with the AST. This guidance is presented in 10 CFR 50.67,
Regulatory Guide 1.183, and Standard Review Plan Section 15.0.1.
Therefore, the proposed amendment does not result in a significant
increase in the consequences or a significant increase in the
probability of any previously evaluated accident.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Implementation of AST does not alter any design basis accident
initiators. These changes do not affect the design function or mode
of operations of systems, structures, or components in the facility
prior to a postulated accident. Since systems, structures, and
components are operated essentially no differently after the AST
implementation, no new failure modes are created by this proposed
change. Therefore, the proposed license amendments will not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The changes proposed are associated with a revision to the
licensing basis for BFN. The results of accident analyses revised in
support of the proposed change are subject to the acceptance
criteria in 10 CFR 50.67. The analyzed events have been carefully
selected, and the analyses supporting this submittal have been
performed using approved methodologies. The dose consequences of
these limiting events are within the acceptance criteria provided by
the regulatory guidance as presented in 10 CFR 50.67, Regulatory
Guide 1.183, and SRP 15.0.1.
Therefore, because the proposed changes continue to result in
dose consequences within the applicable regulatory limits, the
changes are considered to not result in 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 are satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: William F. Burton (Acting).
Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328,
Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: March 3, 2004 (TSC 03-10).
Description of amendment request: The proposed amendment would
revise the Updated Final Safety Analysis Report (UFSAR) and the
Technical Specification Bases description of the seismic qualification
of round flexible ducting, triangular ducting, and associated air bars
installed as part of the suspended ceiling air delivery system in the
main control room.
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?
No. The design function of the MCR [main control room] ducting
system is to support pressurization and cooling of the control room
during normal and accident conditions. The design function of the
MCR suspended ceiling is to remain in place during and subsequent to
an accident, support the triangular and flexible ducts, and not
damage safety-related equipment. The MCR ducting, including the
classification and methodology changes, is a passive feature and
does not act as an accident initiator, i.e., failure of the ducting
would not initiate a design basis accident. The MCR suspended
ceiling has been qualified such that it will remain in place and
perform its safety function during and after an accident.
Consequently, the changes associated with the MCR ducting and
suspended ceiling do not affect the frequency of occurrence for
accidents previously evaluated in the UFSAR.
For the principal design basis accidents, loss of coolant
accident (LOCA), internal flood, steam generator tube rupture
(SGTR), main steam line break (MSLB), etc., the integrity of the MCR
HVAC [heating, ventilation and air conditioning] system, including
the suspended ceiling, will not be compromised. These accidents do
not have a structural effect on the MCR. This means that for
radiological or toxic chemical accidents, the ability to both
pressurize and maintain MCR temperatures within the design limits is
unaffected by the limited quality and seismic requirements for the
flexible and triangular ducting.
An accident that involves a fire that affects the MCR or the
habitability of the MCR was not a consideration for the
qualification of the air distribution components. A fire of this
nature will result in plant operation from the Auxiliary Control
Room (ACR) which is supported by a separate HVAC system.
[[Page 22884]]
The physical effects of an earthquake (including the design
basis SSE) is the only event in which the design basis for the MCR
HVAC is potentially challenged. An evaluation by an industry seismic
expert shows that the ducting and suspended ceiling will remain in
place, will retain their structural integrity such that flow will
not be impeded, and the ducting pressure boundary will not be lost.
Thus, reducing the QA [quality assurance] and seismic qualification
requirements for the MCR ducting and changing the method of seismic
qualification will not result in loss of safety function for any
design basis accident or event. Thus, the accident dose as
previously evaluated in the UFSAR is not affected by the proposed
license amendment.
Based on the above discussion, 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?
No. The MCR ducting addressed by the proposed amendment is not
an accident initiator; i.e., failure of the ducting will not
initiate a design basis accident. In addition, the subject ducting
and suspended ceiling have been evaluated and a determination has
been made that they will continue to perform their safety functions
during normal and accident conditions. Consequently, this activity
does not create a possibility of a new or different type of accident
than any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The changes addressed in TVA' s proposed amendment are
associated with changes in QA requirements and seismic qualification
methodology for safety related air delivery components and for the
suspended ceiling. The change does not affect specific HVAC
equipment safety limits, design limits, set points, or other
critical parameters. In addition, the new seismic analysis
methodology and limited QA requirements ensure that these components
will continue to perform their safety functions during normal and
accident conditions. The previously implied margin of safety against
structural or functional failure of the air delivery components or
suspended ceiling during and after a design basis SSE [safe-shutdown
earthquake] has not been reduced. 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: William F. Burton, Acting.
Tennessee Valley Authority (TVA), Docket No. 50-390, Watts Bar Nuclear
Plant (WBN), Unit 1, Rhea County, Tennessee
Date of amendment request: April 7, 2004.
Description of amendment request: The proposed amendment would
revise the maximum ultimate heat sink (UHS) temperature by revising the
Technical Specification (TS) maximum essential raw cooling water (ERCW)
temperature limit.
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?
No. The proposed change to increase the UHS maximum temperature
will not adversely alter the function, design, or operating
practices for plant systems or components. The UHS is utilized to
remove heat loads from plant systems during normal and accident
conditions. This function is not expected or postulated to result in
the generation of any accident and continues to adequately satisfy
the associated safety functions with the proposed changes.
Therefore, the probability of an accident presently evaluated in the
safety analyses will not be increased. With the exception of re-
gearing the shutdown board room chiller compressors, no other plant
equipment must be altered as a result of this change. Re-gearing of
the shutdown board room chillers will ensure their continued
performance in accordance with design concurrent with the increased
UHS temperature. The heat loads that the UHS is designed to
accommodate have been evaluated for functionality with the higher
temperature limits. The result of these evaluations is that there is
existing margin associated with the systems that utilize the UHS for
normal and accident conditions. These margins are sufficient to
accommodate the postulated normal and accident heat loads with the
proposed changes to the UHS. Since the safety functions of the UHS
are maintained, the systems that ensure acceptable offsite dose
consequences will continue to operate as designed. The change in the
maximum calculated containment pressure associated with the design
basis loss of coolant accident remains below the ASME [American
Society of Mechanical Engineers] Code design internal pressure. The
change to clarify the maximum allowable internal containment
pressure is administrative consistent with present wording in the TS
Bases. Therefore, the consequence of any accident will be the same
as those previously analyzed.
Therefore, since the UHS safety function will continue to meet
accident mitigation requirements and limit dose consequences to
acceptable levels, TVA has concluded that the proposed TS 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?
No. The UHS function provides accident mitigation capabilities
and serves as a heat sink for normal and upset plant conditions; the
UHS is not an initiator of any accident. By allowing the proposed
change in the UHS temperature requirements, only the parameters for
UHS operation are changed while the safety functions of the UHS and
systems that transfer the heat sink capability continue to be
maintained. The proposed change does not impact the response of the
systems and components assumed in the safety analysis. 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?
No. The proposed change has been evaluated for systems that are
needed to support accident mitigation functions as well as normal
operational evolutions. Operational margins were found to exist in
the systems that utilize the UHS capabilities such that these
proposed changes will not result in the loss of any safety function
necessary for normal or accident conditions. The ERCW system has
excess flow margins that will accommodate the increased flows
necessary for the proposed temperature increase. While operating
margins have been reduced by the proposed changes, safety margins
have been maintained as assumed in the accident analyses for
postulated events. The proposed change results in an increase in the
maximum calculated containment peak pressure. However, the change in
the maximum calculated containment peak pressure associated with the
design basis LOCA [loss-of-coolant accident] is a small percentage
of the margin between the current maximum calculated containment
peak pressure and the ASME Code design internal pressure. The change
to clarify the maximum allowable internal containment pressure is
administrative. This aspect of the proposed change does not involve
a significant reduction in a margin of safety. Additionally, the
proposed changes do not require any further modification (the
shutdown board room chiller will be re-geared) of component
setpoints or operating provisions that are necessary to maintain
margins of safety established by the WBN design. 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: General Counsel, Tennessee Valley Authority,
[[Page 22885]]
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: William F. Burton, Acting.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
STP Nuclear Operating Company, Docket No. 50-499, South Texas Project,
Unit 2, Matagordo County, Texas
Date of amendment request: March 4, 2004.
Brief description of amendment request: The proposed amendment
would allow South Texas Project (STP) Unit 2 to change modes with
standby diesel generator 22 inoperable. This is a one-time change that
would expire 14 days after entering Mode 4 on restart from the STP Unit
2 Spring 2002 refueling outage.
Date of publication of individual notice in Federal Register: March
23, 2004.
Expiration date of individual notice: April 22, 2004 (public
comments), and May 24, 2004 (hearing requests).
STP Nuclear Operating Company, Docket No. 50-499, South Texas Project,
Unit 2, Matagordo County, Texas
Date of amendment request: March 18, 2004.
Brief description of amendment request: These amendments revise
Technical Specification (TS) Surveillance Requirement 4.7.7.e.3 to add
a footnote that allows an evaluation for points that do not meet the 1/
8 inch Water Gauge criterion of the current TS. These amendments close
out Notice of Enforcement Discretion No. 04-6-001, which the Commission
granted on March 23, 2004.
Date of publication of individual notice in Federal Register: April
5, 2004.
Expiration date of individual notice: April 19, 2004 (public
comments), and June 4, 2004 (hearing requests).
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: March 23, 2004.
Description of amendments request: To allow both trains of control
room air-conditioning system to be inoperable for up to 7 days,
provided control room temperatures are verified every 4 hours to be
less than or equal to 90 degrees Fahrenheit.
Date of publication of individual notice in the Federal Register:
April 14, 2004 (69 FR 19880).
Expiration date of individual notice: May 14, 2004.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by 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 amendments: July 14, 2003, as supplemented
December 5, 2003, and February 12, 2004.
Brief description of amendments: These amendments change the
Surveillance Requirement 3.6.6.8 to verify each containment spray
nozzle is unobstructed only following maintenance that could result in
nozzle blockage.
Date of issuance: April 8, 2004.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 264 and 241.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 19, 2003 (68 FR
49814). The supplements dated December 5, 2003, and February 12, 2004,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated April 8, 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: February 25, 2004.
Brief description of amendments: These amendments changes the
implementation date for the new cooldown rates for pressure temperature
[[Page 22886]]
limits established by Amendment Nos. 261 and 238 for Calvert Cliffs
Nuclear Power Plant, Unit Nos. 1 and 2, respectively, from 120 days
after issuance, to July 1, 2004.
Date of issuance: April 5, 2004.
Effective date: As of the date of issuance, immediately changing
the implementation date of Amendment Nos. 261 and 238 to July 1, 2004.
Amendment Nos.: 263 and 240.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 5, 2004 (69 FR
10487). The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated April 5, 2004.
No significant hazards consideration comments received: No.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of application for amendment: March 20, 2003; as supplemented
on March 31, April 17, June 11, July 21, and December 11, 2003; and
January 20, February 10, and March 11, 2004.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to reflect an expanded operating domain resulting
from the implementation of the Average Power Range Monitor, Rod Block
Monitor TSs/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA).
Date of Issuance: April 14, 2004.
Effective date: As of the date of issuance, and shall be
implemented at the start of operating cycle 24.
Amendment No.: 219.
Facility Operating License No. DPR-28: Amendment revised the TS.
Date of initial notice in Federal Register: April 15, 2003 (68 FR
18276). The licensee's March 31, April 17, June 11, July 21, and
December 11, 2003; and January 20, February 10, and March 11, 2004,
letters provided clarifying information that did not change the scope
of the proposed amendment as described in the original notice of
proposed action published in the Federal Register, and did not change
the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated April 14, 2004.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi; Entergy Gulf States, Inc., and Entergy Operations,
Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana
Parish, Louisiana; and Entergy Operations, Inc., Docket No. 50-382,
Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of application for amendment: November 6, 2002, as
supplemented by letters dated November 18, 2003, and January 30, 2004.
Brief description of amendment: The amendment would revise the
Facility Operating Licenses, Appendix B, Environmental Protection Plan
(EPP) (Non-Radiological) for the respective plants.
Date of issuance: April 12, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos: 165, Docket No. 50-416, NPF-29; 138, Docket No. 50-
458, NPF-47; 193, Docket No. 50-382, NPF-38.
Facility Operating License Nos. NPF-29, NPF-47, and NPF-38: The
amendments revise the EPPs for the respective plants.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75872).
The licensee enclosed a revised no significant hazards
consideration (NSHC) determination with the supplemental letter dated
November 18, 2003. This revised NSHC determination contained minor
wording changes as compared with the NSHC determination included in the
original application dated November 6, 2002, changes made to reflect
the new EPP changes, and did not expand the scope of the application as
originally noticed, and did not change the conclusions of the NSHC
determination as published in the Federal Register on December 10, 2002
(67 FR 75872). The January 30, 2004, supplemental letter provided
further clarification to the November 18, 2003, supplemental letter
that did not change the conclusion of the NSHC determination published
on December 10, 2002.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 12, 2004.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: June 20, 2003, as supplemented
by letter dated December 12, 2003.
Brief description of amendment: The amendment authorizes changes to
the surveillance requirements for containment integrated leak rate
testing in TS 4.4.a, ``Integrated Leak Rate Tests (Type A).''
Date of issuance: April 6, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 173.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 22, 2003 (68 FR
43391) . The supplemental letter 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 amendment is contained
in a Safety Evaluation dated April 6, 2004.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: March 27, 2003, as supplemented
by letters dated October 30, and December 19, 2003.
Brief description of amendments: The proposed amendment would
approve a selective scope application of an alternative source term for
fuel-handling accidents. Specifically, the amendments would revise
Technical Specification 3.9.3, ``Containment Penetrations,'' to (1)
change the Applicability statement to ``During movement of recently
irradiated fuel assemblies within containment,'' and (2) modify the
Required Action for Condition A to eliminate the requirement to suspend
core alterations and add the requirement to suspend movement of
recently irradiated fuel assemblies within containment if one or more
containment penetrations are not in the required status.
Date of issuance: April 2, 2004.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 213 and 218.
[[Page 22887]]
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 13, 2003 (68 FR
25656). 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 April 2, 2004.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of application for amendment: February 25, 2003, as
supplemented September 9, 2003.
Brief description of amendment: The amendment added an allowed-
outage time for Engineered Safety Features Actuation System
Instrumentation channels to be out of service in a bypassed state.
Date of issuance: April 5, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 167.
Facility Operating License No. NPF-12: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 1, 2003 (68 FR
15762). The September 9, 2003, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination or expand the scope of the application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 5, 2004.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company (STPNOC), Docket Nos. 50-498 and 50-499,
South Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: March 18, 2004, as supplemented by
letters dated April 7 and 13, 2004.
Brief description of amendments: The amendments revise TS
Surveillance Requirement (SR) 4.7.7.e.3 to add a footnote that allows
use of alternate criteria for those measured points at positive
pressure but that do not meet the \1/8\ inch Water Gauge criterion of
the current TS. In addition the word ``that'' in the second line of the
original text of SR 4.7.7.e.3 is changed to ``than'' to correct an
existing typographical error. These amendments supersede Notice of
Enforcement Discretion (NOED) No. 04-6-001, which the Commission staff
granted to STPNOC on March 23, 2004.
Date of issuance: April 15, 2004.
Effective date: As of the date of issuance.
Amendment Nos.: Unit 1-161; Unit 2-151.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC):
Yes. A notice was published in the Federal Register on April 5,
2004 (69 FR 17718). The notice provided an opportunity to submit
comments on the Commission's proposed NSHC determination. No comments
have been received. The notice also provided an opportunity to request
a hearing within 60 days from the date of publication, but indicated
that if the Commission makes a final NSHC determination, any such
hearing would take place after issuance of the amendment. The
supplements dated April 7 and 13, 2004, provided additional information
that clarified the application, did not expand the scope of the
application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 15, 2004.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: April 11, 2003, as supplemented
by the October 2, 2003, meeting, and a letter dated February 20, 2004.
Description of amendment request: The amendments revised Technical
Specification (TS) Table 3.3.5.1-1 which will result in a change to the
Updated Final Safety Analysis Report (UFSAR), Table 6.5-3.
Date of issuance: April 1, 2004.
Effective date: Date of issuance, to be implemented within 60 days
for Unit 1, during Cycle 13 Refueling Outage for Unit 2 , and during
Cycle 12 Refueling Outage for Unit 3.
Amendment Nos.: 250, 289 & 248.
Facility Operating License Nos. DPR-33, DPR-52, and DPR-68:
Amendments revised the TSs which will result in a change the UFSAR,
Table 6.5-3.
Date of initial notice in Federal Register: May 27, 2003 (68 FR
28857). The October 2, 2003, meeting, and the February 20, 2004,
letter, provided clarifying information that did not change the scope
of the original request or the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 1, 2004.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: June 27, 2003, as supplemented
by letters dated December 9, 2003, January 14 and April 5, 2004.
Brief description of amendment: The amendment approves the
application of leak-before-break methodology for the accumulator and
residual heat removal lines and installation of an opening in the
secondary shield wall in terms of the effect of the opening on
occupational exposure. The shield wall opening is related to plant
modifications that would facilitate maintenance on the replacement
steam generators to be installed in Refueling Outage 14 (Fall 2005).
Date of issuance: April 12, 2004.
Effective date: April 12, 2004, and shall be implemented prior to
entering Mode 4 during the startup from Refueling Outage 13 which is
scheduled for the Spring of 2004.
Amendment No.: 161.
Facility Operating License No. NPF-30: The amendment revised the
Final Safety Analysis Report.
Date of initial notice in Federal Register: July 22, 2003 (68 FR
43397).
The December 9, 2003, January 14 and April 5, 2004, supplemental
letters provided additional clarifying information, did not expand the
scope of the application as originally noticed, and did not change the
staff's original proposed no significant hazards determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 12, 2004.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket No. 50-339, North Anna
Power Station, Unit 2, Louisa County, Virginia
Date of application for amendment: March 28, 2002, as supplemented
by letters dated May 13, June 19, July 9, July 25, August 2, August 16,
and November 15, 2002, May 6, May 9, May 27, June 11 (2 letters), July
18, August 20, August 26, September 4, September 5, September 22,
September 26 (2
[[Page 22888]]
letters), November 10, December 8, and December 17, 2003, and January
6, January 22 (2 letters), February 12, February 13, and March 1, 2004.
The November 15, 2002, submittal replaced the submittals dated July 9,
July 25, and August 16, 2002.
Brief description of amendment: This amendment revises Improved
Technical Specification Sections 2.1, 4.2, and 5.6.5 in order to allow
Virginia Electric and Power Company to implement Framatome ANP Advanced
Mark-BW fuel at North Anna Power Station, Unit 2.
Date of issuance: April 1, 2004.
Effective date: As of the date of issuance and shall be implemented
prior to the initiation of core onload during Refueling Outage 16
(Spring 2004).
Amendment No.: 216.
Renewed Facility Operating License No. NPF-7: Amendment changes the
Improved Technical Specifications.
Date of initial notice in Federal Register: July 22, 2003 (68 FR
43397). The supplements dated July 18, August 20, August 26, September
4, September 5, September 22, September 26 (2 letters), November 10,
December 8, and December 17, 2003, and January 6, January 22 (2
letters), February 12, February 13, and March 1, 2004, contained
clarifying information only and did not change the initial no
significant hazards consideration determination or expand the scope of
the initial application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 1, 2004.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22.
Therefore, pursuant to 10 CFR 51.22(b), no environmental impact
statement or environmental assessment need be prepared for these
amendments. If the Commission has prepared an environmental assessment
under the special circumstances provision in 10 CFR 51.12(b) and has
made a determination based on that assessment, it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland. Publicly available
records will be accessible from the Agencywide Documents Access and
Management System's (ADAMS) Public Electronic Reading Room on the
Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html.
If you do not have access to ADAMS or if there are problems in
accessing the documents located in ADAMS, contact the NRC Public
Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or
by e-mail to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, the licensee may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene. Requests for a hearing and a petition for leave
to intervene shall be filed in accordance with the Commission's ``Rules
of Practice for Domestic Licensing Proceedings'' in 10 CFR part 2.
Interested persons should consult a current copy of 10 CFR 2.309, which
is available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1-
[[Page 22889]]
800-397-4209, 301-415-4737, or by e-mail to [email protected]. If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
---------------------------------------------------------------------------
\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) e-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to 301-415-3725 or by email to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer or the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
Pacific Gas and Electric Company, Docket No. 50-275, Diablo Canyon
Power Plant, Unit No. 1, San Luis Obispo County, California
Date of application for amendment: April 2, 2004, as superseded by
application dated April 8, 2004.
Description of amendment request: The amendment revises the
Technical Specification 3.3.5, ``Loss of Power (LOP) Diesel Generator
(DG) Start Instrumentation,'' to allow performance of Surveillance
Requirement (SR) 3.3.5.2 for the trip actuation device operational
test, prior to first entry into MODE 4, by adding a note to the
FREQUENCY column of SR 3.3.5.2 on a one-time basis.
Date of issuance: April 15, 2004.
Effective date: April 15, 2004, and shall be implemented within 10
days from the date of issuance.
Amendment No.: 165.
Facility Operating License No. DPR-80: The amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. A public notice was published in the San
Luis Obispo Tribune on April 13 and 14, 2004. The notice provided an
opportunity to submit comments on the Commission's proposed NSHC
determination. No comments have been received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and
[[Page 22890]]
final NSHC determination are contained in a safety evaluation dated
April 15, 2004.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
Dated in Rockville, Maryland, this 19th day of April, 2004.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 04-9225 Filed 4-26-04; 8:45 am]
BILLING CODE 7590-01-P