[Federal Register Volume 76, Number 6 (Monday, January 10, 2011)]
[Notices]
[Pages 1462-1469]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2011-215]
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NUCLEAR REGULATORY COMMISSION
[NRC-2010-0390]
Notice Applications and Amendments to Facility Operating Licenses
Involving Proposed No Significant Hazards Considerations and Containing
Sensitive Unclassified Non-Safeguards Information and Order Imposing
Procedures for Access to Sensitive Unclassified Non-Safeguards
Information
I. Background
Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC) is publishing this 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 notice includes notices of amendments containing sensitive
unclassified non-safeguards information (SUNSI).
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 Title 10 of the Code of Federal
Regulations (10 CFR), Section 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.
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,
Announcements and Directives Branch (RADB), TWB-05-B01M, 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 faxed to the RADB at 301-492-3446.
Documents may be examined, and/or copied for a fee, at the NRC's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1 F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license.
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 person(s)
should consult a current copy of 10 CFR 2.309, which is available at
the Commission's PDR, located at One White Flint North, Public File
Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland, or
at http://www.nrc.gov/reading-rm/doc-collections/cfr/part002/part002-0309.html. Publicly available records will be accessible from the
Agencywide Documents Access and Management System's (ADAMS) Public
Electronic
[[Page 1463]]
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm.html. 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 requestor/petitioner
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
requestor/petitioner 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 requestor/
petitioner intends to rely in proving the contention at the hearing.
The requestor/petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
requestor/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. 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 requestor/petitioner to relief. A requestor/
petitioner 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.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC E-Filing rule (72 FR 49139,
August 28, 2007). The E-Filing process requires participants to submit
and serve all adjudicatory documents over the Internet, or in some
cases to mail copies on electronic storage media. Participants may not
submit paper copies of their filings unless they seek an exemption in
accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
ten (10) days prior to the filing deadline, the participant should
contact the Office of the Secretary by e-mail at
[email protected], or by telephone at (301) 415-1677, to request
(1) a digital ID certificate, which allows the participant (or its
counsel or representative) to digitally sign documents and access the
E-Submittal server for any proceeding in which it is participating; and
(2) advise the Secretary that the participant will be submitting a
request or petition for hearing (even in instances in which the
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the
Secretary will establish an electronic docket for the hearing in this
proceeding if the Secretary has not already established an electronic
docket.
Information about applying for a digital ID certificate is
available on NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing
the E-Submittal server are detailed in NRC's ``Guidance for Electronic
Submission,'' which is available on the agency's public Web site at
http://www.nrc.gov/site-help/e-submittals.html. Participants may
attempt to use other software not listed on the Web site, but should
note that the NRC's E-Filing system does not support unlisted software,
and the NRC Meta System Help Desk will not be able to offer assistance
in using unlisted software.
If a participant is electronically submitting a document to the NRC
in accordance with the E-Filing rule, the participant must file the
document using the NRC's online, Web-based submission form. In order to
serve documents through EIE, users will be required to install a Web
browser plug-in from the NRC Web site. Further information on the Web-
based submission form, including the installation of the Web browser
plug-in, is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a
docket has been created, the participant can then submit a request for
hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with NRC guidance
available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the
documents are submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
E-Filing system also distributes an e-mail notice that provides access
to the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or
[[Page 1464]]
their counsel or representative) must apply for and receive a digital
ID certificate before a hearing request/petition to intervene is filed
so that they can obtain access to the document via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC Web site
at http://www.nrc.gov/site-help/e-submittals.html, by e-mail at
[email protected], or by a toll-free call at (866) 672-7640. The
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m.,
Eastern Time, Monday through Friday, excluding government holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted 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; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, Sixteenth
Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland
20852, Attention: Rulemaking and Adjudications Staff. Participants
filing a document in this manner are responsible for serving the
document on all other participants. Filing is considered complete by
first-class mail as of the time of deposit in the mail, or by courier,
express mail, or expedited delivery service upon depositing the
document with the provider of the service. A presiding officer, having
granted an exemption request from using E-Filing, may require a
participant or party to use E-Filing if the presiding officer
subsequently determines that the reason for granting the exemption from
use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant
to an order of the Commission, or the presiding officer. Participants
are requested not to include personal privacy information, such as
Social Security numbers, home addresses, or home phone numbers in their
filings, unless an NRC regulation or other law requires submission of
such information. With respect to copyrighted works, except for limited
excerpts that serve the purpose of the adjudicatory filings and would
constitute a Fair Use application, participants are requested not to
include copyrighted materials in their submission.
Petitions for leave to intervene must be filed no later than 60
days from the date of publication of this notice. Non-timely filings
will not be entertained absent a determination by the presiding officer
that the petition or request should be granted or the contentions
should be admitted, based on a balancing of the factors specified in 10
CFR 2.309(c)(1)(i)-(viii).
For further details with respect to this amendment 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 electronically from the
ADAMS Public Electronic Reading Room on the Internet at the NRC Web
site, http://www.nrc.gov/reading-rm/adams.html. If you do not have
access to ADAMS or if there are problems in accessing the documents
located in ADAMS, contact the PDR Reference staff at 1-800-397-4209,
301-415-4737, or by e-mail to [email protected].
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: September 8, 2010, as supplemented by
letters dated November 18 and 23, 2010.
Description of amendment request: This amendment request contains
sensitive unclassified non-safeguards information (SUNSI). The proposed
license amendment request will increase the maximum reactor core power
operating limit from 3,898 megawatts thermal (MWt) to 4,408 MWt at
Grand Gulf Nuclear Station (GGNS), Unit 1. The following Operating
License (OL) and Technical Specification (TS) sections, and associated
TS bases, will be revised as a result of the proposed extended power
uprate (EPU):
OL Paragraph 2.C.(1) and the addition of new license
conditions
Definitions--Rated Thermal Power (RTP) and a new
definition for Pressure and Temperature Limits Report (PTLR)
Thermal Power Limit with Low Dome Pressure or Low Core
Flow (TS 2.1.1.1)
Minimum Critical Power Ratio (MCPR) Safety Limit (TS
2.1.1.2)
Standby Liquid Control (SLC) System (TS 3.1.7)
Average Planar Linear Heat Generation Rate (APLHGR) (TS
3.2.1)
Minimum Critical Power Ratio (MCPR) (TS 3.2.2)
Linear Heat Generation Rate (LHGR) (TS 3.2.3)
Reactor Protection System (RPS) Instrumentation (TS
3.3.1.1)
End of Cycle Recirculation Pump Trip (EOC-RPT)
Instrumentation (TS 3.3.4.1)
Primary Containment and Drywell Isolation Instrumentation
(TS 3.3.6.1)
Jet Pumps (TS 3.4.3)
Safety/Relief Valves (TS 3.4.4)
Reactor Coolant System (RCS) Pressure and Temperature (P/
T) Limits (TS 3.4.11)
Main Turbine Bypass System (New TS 3.7.7), and
RCS Pressure and Temperature Limits Report (PTLR) (New TS
5.6.6).
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 increase in power level does not significantly
increase the probability or consequences of an accident previously
evaluated.
The proposed change will increase the maximum authorized core
power level for GGNS from the current licensed thermal power (CLTP)
of 3,898 megawatts thermal (MWt) to 4,408 MWt. Evaluations and
analyses of the nuclear steam supply system (NSSS) and balance of
plant (BOP) structures, systems, and components (SSCs) that could be
affected by the power uprate were performed in accordance with the
approaches described in:
NEDC-33004P-A (commonly called CLTR), Licensing Topical
Report Constant Pressure Power Uprate, Revision 4;
NEDC-32424P-A (commonly called ELTR1), Generic
Guidelines for General Electric Boiling Water Reactor Extended Power
Uprate; and
NEDC-32523P-A (commonly called ELTR2), Generic
Evaluations of General Electric Boiling Water Reactor Extended Power
Uprate.
The evaluations concluded that all plant components, as
modified, will continue to be capable of performing their design
function at the proposed uprated core power level.
The GGNS licensing and design bases, including GGNS accident
analyses, were also evaluated for the effect of the proposed power
increase. The evaluation concluded that the applicable analysis
acceptance criteria continue to be met. Power level is not an
initiator of any transient or accident; it is
[[Page 1465]]
used as an input assumption to equipment design and accident
analyses.
The proposed change does not affect the release paths or the
frequency of release for any accidents previously evaluated in the
[Updated Final Safety Analysis Report]. Structures, systems, and
components required to mitigate transients remain capable of
performing their design functions considering radiological
consequences associated with the effect of the proposed EPU. The
source terms used to evaluate the radiological consequences were
reviewed and were determined to bound operation at EPU power levels.
The results of EPU accident evaluations do not exceed NRC-approved
acceptance limits.
The spectrum of postulated accidents and transients were
reviewed and were shown to meet the regulatory criteria to which
GGNS is currently licensed. In the area of fuel and core design, the
Safety Limit Minimum Critical Power Ratio (SLMCPR) and other
Specified Acceptable Fuel Design Limits (SAFDLs) are still met.
Continued compliance with the [SLMCPR] and other SAFDLs is confirmed
on a cycle specific basis consistent with the criteria accepted by
the NRC.
Challenges to the reactor coolant pressure boundary were
evaluated at EPU conditions (pressure, temperature, flow, and
radiation) and found to meet the acceptance criteria for allowable
stresses. Adequate overpressure margin is maintained.
Challenges to the containment were also evaluated. Containment
and its associated cooling system continue to meet applicable
regulatory requirements. The increase in the calculated post Loss of
Coolant Accident (LOCA) suppression pool temperature above the
current design limit was evaluated and determined to be acceptable.
Radiological releases were evaluated and found to be within the
regulatory limits of 10 CFR 50.67, Accident Source Terms.
Change in Methodologies
The use of more accurate modeling of the annulus pressurization
loads is not relevant to accident initiation, but rather, pertains
to the method used to accurately evaluate annulus pressurization
during postulated accidents. The use of a new method does not, in
any way, alter any fission product barrier or SSC and provides a
better representation of dynamic behavior.
The GGNS containment analysis was performed using the SHEX
computer code, which is not relevant to accident initiation.
The GGNS steam dryer evaluation was performed using a plant
based load evaluation method. The use of this evaluation is not
relevant to accident initiation. The steam dryer is a non-safety
related component.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No, the increase in power does not create the
possibility of a new or different kind of accident from any
previously evaluated.
The proposed change increases the maximum authorized core power
level for GGNS from the CLTP of 3898 MWt to 4408 MWt. An evaluation
of the equipment that could be affected by the power uprate has been
performed. No new operating modes, safety-related equipment lineups,
accident scenarios, or equipment failure modes were identified. The
full spectrum of accident considerations was evaluated and no new or
different kinds of accidents were identified. For GGNS, the standard
evaluation methods outlined in CLTR, ELTR1, and ELTR2 were applied
to the capability of existing or modified safety-related plant
equipment. No new accidents or event precursors were identified.
All SSCs previously required for the mitigation of a transient
remain capable of fulfilling their intended design functions. The
proposed increase in power does not adversely affect safety-related
systems or components and does not challenge the performance or
integrity of any safety-related system. The change does not
adversely affect any current system interfaces or create any new
interfaces that could result in an accident or malfunction of a
different kind than was previously evaluated. Operating at the
proposed EPU power level does not create any new accident initiators
or precursors.
Change in Methodologies
The use of more accurate modeling of the annulus pressurization
loads is not relevant to accident initiation, but rather, pertains
to the method used to accurately evaluate annulus pressurization
during postulated accidents. The use of this methodology does not
involve any physical changes to plant structures or systems, and
does not create a new initiating event for the spectrum of events
currently postulated. Further, the methodologies do not result in
the need to postulate any new accident scenarios.
The GGNS containment analysis was performed using the SHEX
computer code, which is not an accident initiator and therefore does
not result in the creation of any new accidents.
The use of the plant based load evaluation method to perform the
GGNS steam dryer analysis does not result in the creation of any new
accidents since the steam dryer is not safety-related and is not
considered an accident initiator.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No, the proposed increase in power does not involve a
significant reduction in a margin of safety.
Based on the analyses of the proposed power increase, the
relevant design and safety acceptance criteria will be met without a
significant reduction in margins of safety. The analyses supporting
EPU have demonstrated that the GGNS SSCs are capable of safely
performing at EPU conditions. The analyses identified and defined
the major input parameters to the NSSS, analyzed NSSS design
transients, and evaluated the capabilities of the NSSS fluid
systems, NSSS/BOP interfaces, NSSS control systems, and NSSS and BOP
components, as appropriate. Radiological consequences of design
basis events remain within regulatory limits and are not increased
significantly. The analyses confirmed that NSSS and BOP SSCs are
capable, some with modifications, of achieving EPU conditions
without significant reduction in margins of safety.
Analyses have shown that the integrity of primary fission
product barriers will not be significantly affected as a result of
the power increase. Calculated loads on SSCs important to safety
have been shown to remain within design allowable under EPU
conditions for all design basis event categories. Plant response to
transients and accidents do not result in exceeding acceptance
criteria.
As appropriate, the evaluations that demonstrate acceptability
of EPU have been performed using methods that have either been
reviewed and approved by the NRC staff, or that are in compliance
with regulatory review guidance and standards established for
maintaining adequate margins of safety. These evaluations
demonstrate that there are no significant reductions in the margins
of safety.
Maximum power level is one of the inherent inputs that determine
the safe operating range defined by the accident analyses. The
Technical Specifications ensure that GGNS is operated within the
bounds of the inputs and assumptions used in the accident analyses.
The acceptance criteria for the accident analyses are conservative
with respect to the operating conditions defined by the Technical
Specifications. The engineering reviews performed for the constant
pressure extended power uprate confirm that the accident analyses
criteria are met at the revised maximum allowable thermal power
level of 4408 MWt. Therefore, the adequacy of the revised Facility
Operating License and Technical Specifications to maintain the plant
in a safe operating range is also confirmed, and the increase in
maximum allowable power level does not involve a significant
decrease in a margin of safety.
Change in Methodologies
The use of more accurate modeling of the annulus pressurization
loads is not relevant to accident initiation, but rather, pertains
to the method used to accurately evaluate annulus pressurization
during postulated accidents. The use of a more accurate methodology
to generate mass and energy release rates reduces the potential for
methodology induced response profile frequency shifts that could
result in a non-conservative load assessment. The use of more
accurate methods, to minimize the impact of methodology induced
response profile frequency shifts, does not result in a reduction in
the margin of safety.
In light of issues identified in GEH [GE-Hitachi Nuclear Energy
Americas LLC] Safety Information Concern SC 09-01, Annulus
Pressurization Loads Evaluation, dated June 8, 2009, a realistic
annulus pressurization methodology is required to ensure that the
[[Page 1466]]
frequency content of the annulus pressurization transient is
captured and correctly accounted for in the downstream structural,
component and piping load analyses. The use of more accurate
modeling of the annulus pressurization loads does not adversely
impact containment SSCs or the subcompartments.
The GGNS containment analysis was performed using the SHEX
computer code. The results of the containment analysis demonstrate
that the containment remains within all of its design limits
following the most limiting design basis accident.
The steam dryer evaluation was performed in accordance with
[NRC] Regulatory Guide 1.20, Comprehensive Vibration Assessment
Program for Reactor Internals During Preoperational and Initial
Startup Testing. The non-safety related replacement steam dryer
conservatively exceeds the vibration and stress requirements.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Joseph A. Aluise, Associate General
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: September 22, 2010.
Description of amendment request: This amendment request contains
sensitive unclassified non-safeguards information (SUNSI). The proposed
amendment would modify the Facility Operating License and Technical
Specifications (TSs) to allow Hope Creek Generating Station (HCGS) to
operate at a reduced feedwater temperature for purposes of extending
the normal fuel cycle. The amendment would also allow operation with
feedwater heaters out-of-service at any time during the operating
cycle.
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 with Nuclear Regulatory
Commission (NRC) staff edits in square brackets:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The effect of FWTR [feedwater temperature reduction] on the
probability and consequences of accidents, Anticipated Operational
Occurrences (AOO) and events in the Updated Final Safety Analysis
(UFSAR) were reviewed.
The impact of FWTR on the Design Basis Accident (DBA) Loss of
Coolant Accident (LOCA) was considered. Evaluations and analyses
were performed to determine that the current Licensing Basis PCT
[peak cladding temperature] remains applicable for operation of HCGS
with FWTR. The calculated maximum fuel element cladding temperature
does not exceed 2,200 [deg]F, the calculated total local oxidation
does not exceed 17% times the total cladding thickness, the
calculated total amount of hydrogen generated from a chemical
reaction of the cladding with water or steam is less than 1% times
the hypothetical amount if all the metal in the cladding cylinder
were to react, the core remains amenable to long term cooling, and
there is sufficient long term core cooling available. Analysis also
demonstrated that FWTR operation at HCGS continues to meet design
limits for the DBA-LOCA peak drywell pressure and temperature.
Therefore, there is no increase in the consequence of an accident
previously evaluated in the UFSAR.
The only AOO that requires consideration in assessing the effect
of FWTR on event consequences is the feedwater controller failure--
increasing flow (FWCF). This is based upon the finding that the
other AOOs are less sensitive to a reduction in feedwater
temperature. The rated power and off-rated Power Distribution
Limits, Critical Power Ratio [CPR] and Linear Heat Generation Rate
[LHGR], for the FWCF event are validated on a cycle specific basis
to ensure compliance with the Safety Limit Minimum Critical Power
Ratio (SLMCPR) and compliance with the fuel rod thermal mechanical
acceptance criteria of avoiding fuel centerline melt and 1% cladding
plastic strain. Consequently, there is no increase in the
consequences of an AOO previously evaluated.
The impact of FWTR on the consequences of the following events
was also considered: Anticipated Transient Without Scram (ATWS),
vessel overpressure, thermal-hydraulic stability, and High Energy
Line Break (HELB). The evaluation of ATWS and vessel overpressure
concluded that the consequences of the events at normal feedwater
temperature remain bounding for FWTR. The evaluation of HELB
determined the impact was bounded by the current design basis.
Thermal-hydraulic stability considerations, as impacted by FWTR,
involve both the determination of a cycle specific OPRM [oscillation
power range monitor] setpoint and determination of a cycle specific
backup stability protection (BSP) regions and corresponding adequacy
of the OPRM trip enabled region. The cycle specific determinations
and validations performed in accordance with NRC-approved methods
ensure that the SLMCPR will be protected if a thermal hydraulic
stability event were to occur. Therefore, there is no increase in
the consequence of these events previously evaluated in the UFSAR.
In addition, the following areas were also evaluated. The
reactor power level and operating pressure are not changed. FWTR has
no effect on the decay heat. Current design limits associated with
long-term containment analyses, including RSLB [recirculation
suction line break], loss of offsite power (LOOP), intermediate
break accident (IBA), small break accident (SBA), and NUREG-0783
safety relief valve (SRV) steam discharge events continue to be
supported without change. Therefore, there is no increase in the
consequence of these events previously evaluated in the UFSAR.
The probability of an accident is not affected by the proposed
changes since no structures, systems or components (SSC) which could
initiate an accident are affected. Therefore, the proposed changes
do not significantly increase the probability of any previously
evaluated accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not alter the design function of any
SSC. The implementation of FWTR operation does not create the
possibility of a new or different kind of accident. Power
Distribution Limits on CPR, LHGR and APLHGR [average planar linear
heat generation rate], and OPRM setpoints, which are determined in
accordance with NRC-approved methods and are included in the Core
Operating Limits Report (COLR), as part of the normal reload
licensing process will continue to assure that core operation is in
accordance with the conditions currently assumed for event
initiation. FWTR was reviewed against the accidents, AOOs and events
in the UFSAR and it was determined there would be no adverse impact;
the existing design basis remains bounding. In addition, the
proposed changes do not involve new system interactions or equipment
modifications to the plant. FWTR does not involve any new type of
testing or maintenance. Therefore there are no new design basis
failure mechanisms, malfunctions, or accident initiators created by
the proposed changes.
The existing low power scram bypass setpoint, based on turbine
first stage pressure and the calculated change in steam flow was
evaluated. At a reduced feedwater temperature, it was concluded that
the reactor scram bypass setting for turbine first stage pressure
was not sufficiently conservative relative to the TS value of 24%
rated thermal power. Therefore a new setpoint of approximately 21.4%
has been calculated. The new set-point increases the low power
bypass set-point conservatism at normal feedwater temperature (NFWT)
and maintains the same conservatism at FFWTR [final feedwater
temperature reduction] conditions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
[[Page 1467]]
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The AOOs and accidents described in the UFSAR were evaluated for
effects caused by the reduced feedwater temperature. For cycle
independent considerations, the evaluations determined that the
consequences of the events are either bounded by the current design
and licensing basis results, are within design acceptance criteria,
or will not change in a manner that would reduce the margin of
safety. For cycle specific considerations, cycle specific analyses
utilizing NRC-approved methods that produce the values of the limits
documented in the COLR will continue to assure that core operation
is maintained within the existing design basis and safety limits. No
design basis or safety limit is altered by the proposed change.
The existing low power scram bypass setpoint, based on turbine
first stage pressure and the calculated change in steam flow was
evaluated. At a reduced feedwater temperature, it was concluded that
the reactor scram bypass setting for turbine first stage pressure
was not sufficiently conservative relative to the TS value of 24%
rated thermal power. Therefore a new setpoint of approximately 21.4%
has been calculated. The new set-point increases the low power
bypass set-point conservatism at NFWT and maintains the same
conservatism at FFWTR conditions.
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, and with the changes noted above in square brackets, 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: Vincent Zabielski, PSEG Nuclear LLC--N21,
P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
Tennessee Valley Authority, Docket Nos. 50-259, Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of amendment request: October 23, 2009, as supplemented by
letters dated November 17, 2009, and April 16, 2010 (TS-473).
Description of amendment request: This amendment request contains
sensitive unclassified non-safeguards information (SUNSI). Tennessee
Valley Authority (the licensee) plans to transition Browns Ferry
Nuclear Plant (BFN), Unit 1 to AREVA fuel. To support the transition,
the proposed amendment adds the AREVA NP analysis methodologies to the
list of approved methods to be used in determining the core operating
limits in the core operating limits report. Additional technical
specification (TS) changes are requested to reflect the AREVA NP
specific methods for monitoring and enforcing the thermal limits. The
licensee request is for nonextended power uprate conditions (i.e., 105
percent of Original Licensed Thermal Power level) only.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1: Does the proposed amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response: No.
Changing fuel designs and making an editorial change to TS will
not increase the probability of a loss of coolant accident. The fuel
cannot increase the probability of a primary coolant system breach
or rupture, as there is no interaction between the fuel and the
system piping. The fuel will continue to meet the 10 CFR 50.46
limits for peak clad temperature, oxidation fraction, and hydrogen
generation. Therefore, the consequences of a LOCA [loss-of-coolant-
accident] will not be increased.
Similarly, changing the fuel design and making an editorial
change to TS cannot increase the probability of an abnormal
operating occurrence (AOO). As a passive component, the fuel does
not interact with plant operating or control systems. Therefore, the
fuel change cannot affect the initiators of the previously evaluated
AOO transient events. Thermal limits for the new fuel will be
determined on a reload specific basis, ensuring the specified
acceptable fuel design limits continue to be met. Therefore, the
consequences of a previously evaluated AOO will not increase.
The refueling accident is potentially affected by a change in
fuel design due to the mechanical interaction between the fuel and
the refueling equipment. However, the probability of the refueling
accident with ATRIUM-10 fuel is not increased because the upper bail
handle is designed to be mechanically compatible with existing fuel
handling equipment. The design weight of the ATRIUM-10 design is
similar to other designs in use at BFN and is well within the design
capability of the refueling equipment. The consequences of the
refueling accident are similar to the current GE14 fuel, remaining
well within the design basis (7x7 Fuel) evaluation in the UFSAR
[Updated Final Safety Analysis Report].
The probability of a control rod drop accident does not increase
because the ATRIUM-10 fuel channel is mechanically compatible with
the co-resident fuel and existing control blade designs. The
mechanical interaction and friction forces between the ATRIUM-10
channel and control blades would not be higher than previous
designs. In addition, routine plant testing includes confirmation of
adequate control blade to control rod drive coupling. The
probability of a rod drop accident is not increased with the use of
ATRIUM-10 fuel. Control rod drop accident consequences are evaluated
on a cycle specific basis, confirming the number of calculated rod
failures remains with the UFSAR design basis.
The dose consequences of all the previously evaluated UFSAR
accidents remain with the limits of 10 CFR 50.67.
Criterion 2: Does the proposed amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response: No.
The ATRIUM-10 fuel product has been designed to maintain
neutronic, thermal-hydraulic, and mechanical compatibility with the
NSSS [Nuclear Steam Supply System] vendor fuel designs. The ATRIUM-
10 fuel has been designed to meet fuel licensing criteria specified
in NUREG-0800, ``Standard Review Plan for Review of Safety Analysis
Reports for Nuclear Power Plants.'' Compliance with these criteria
ensures the fuel will not fail in an unexpected manner. A change in
fuel design and an editorial change to TS cannot create any new
accident initiators because the fuel is a passive component having
no direct influence on the performance of operating plant systems
and equipment. Hence, a fuel design change cannot create a new type
of malfunction leading to a new or different kind of transient or
accident. Consequently, the proposed fuel design change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
Criterion 3: Does the proposed amendment involve a significant
reduction in a margin of safety?
Response: No.
The ATRIUM-10 fuel is designed to comply with the fuel licensing
criteria specified in NUREG-0800. Reload specific and cycle
independent safety analyses are performed ensuring no fuel failures
will occur as the result of abnormal operational transients, and
dose consequences for accidents remain with the bounds of 10 CFR
50.67. All regulatory margins and requirements are maintained.
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 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, 6A West Tower, Knoxville, Tennessee 37902.
NRC Branch Chief: Douglas A. Broaddus.
[[Page 1468]]
Order Imposing Procedures for Access to Sensitive Unclassified Non-
Safeguards Information for Contention Preparation
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
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Tennessee Valley Authority, Docket Nos. 50-259, Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
A. This Order contains instructions regarding how potential parties
to this proceeding may request access to documents containing Sensitive
Unclassified Non-Safeguards Information (SUNSI).
B. Within 10 days after publication of this notice of hearing and
opportunity to petition for leave to intervene, any potential party who
believes access to SUNSI is necessary to respond to this notice may
request such access. A ``potential party'' is any person who intends to
participate as a party by demonstrating standing and filing an
admissible contention under 10 CFR 2.309. Requests for access to SUNSI
submitted later than 10 days after publication will not be considered
absent a showing of good cause for the late filing, addressing why the
request could not have been filed earlier.
C. The requestor shall submit a letter requesting permission to
access SUNSI to the Office of the Secretary, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, and provide a copy to the Associate General
Counsel for Hearings, Enforcement and Administration, Office of the
General Counsel, Washington, DC 20555-0001. The expedited delivery or
courier mail address for both offices is: U.S. Nuclear Regulatory
Commission, 11555 Rockville Pike, Rockville, Maryland 20852. The e-mail
addresses for the Office of the Secretary and the Office of the General
Counsel are [email protected] and [email protected],
respectively.\1\ The request must include the following information:
---------------------------------------------------------------------------
\1\ While a request for hearing or petition to intervene in this
proceeding must comply with the filing requirements of the NRC's
``E-Filing Rule,'' the initial request to access SUNSI under these
procedures should be submitted as described in this paragraph.
---------------------------------------------------------------------------
(1) A description of the licensing action with a citation to this
Federal Register notice;
(2) The name and address of the potential party and a description
of the potential party's particularized interest that could be harmed
by the action identified in C.(1);
(3) The identity of the individual or entity requesting access to
SUNSI and the requestor's basis for the need for the information in
order to meaningfully participate in this adjudicatory proceeding. In
particular, the request must explain why publicly-available versions of
the information requested would not be sufficient to provide the basis
and specificity for a proffered contention;
D. Based on an evaluation of the information submitted under
paragraph C.(3) the NRC staff will determine within 10 days of receipt
of the request whether:
(1) There is a reasonable basis to believe the petitioner is likely
to establish standing to participate in this NRC proceeding; and
(2) The requestor has established a legitimate need for access to
SUNSI.
E. If the NRC staff determines that the requestor satisfies both
D.(1) and D.(2) above, the NRC staff will notify the requestor in
writing that access to SUNSI has been granted. The written notification
will contain instructions on how the requestor may obtain copies of the
requested documents, and any other conditions that may apply to access
to those documents. These conditions may include, but are not limited
to, the signing of a Non-Disclosure Agreement or Affidavit, or
Protective Order \2\ setting forth terms and conditions to prevent the
unauthorized or inadvertent disclosure of SUNSI by each individual who
will be granted access to SUNSI.
---------------------------------------------------------------------------
\2\ Any motion for Protective Order or draft Non-Disclosure
Affidavit or Agreement for SUNSI must be filed with the presiding
officer or the Chief Administrative Judge if the presiding officer
has not yet been designated, within 30 days of the deadline for the
receipt of the written access request.
---------------------------------------------------------------------------
F. Filing of Contentions. Any contentions in these proceedings that
are based upon the information received as a result of the request made
for SUNSI must be filed by the requestor no later than 25 days after
the requestor is granted access to that information. However, if more
than 25 days remain between the date the petitioner is granted access
to the information and the deadline for filing all other contentions
(as established in the notice of hearing or opportunity for hearing),
the petitioner may file its SUNSI contentions by that later deadline.
G. Review of Denials of Access
(1) If the request for access to SUNSI is denied by the NRC staff
either after a determination on standing and need for access, or after
a determination on trustworthiness and reliability, the NRC staff shall
immediately notify the requestor in writing, briefly stating the reason
or reasons for the denial.
(2) The requestor may challenge the NRC staff's adverse
determination by filing a challenge within 5 days of receipt of that
determination with: (a) The presiding officer designated in this
proceeding; (b) if no presiding officer has been appointed, the Chief
Administrative Judge, or if he or she is unavailable, another
administrative judge, or an administrative law judge with jurisdiction
pursuant to 10 CFR 2.318(a); or (c) if another officer has been
designated to rule on information access issues, with that officer.
H. Review of Grants of Access. A party other than the requestor may
challenge an NRC staff determination granting access to SUNSI whose
release would harm that party's interest independent of the proceeding.
Such a challenge must be filed with the Chief Administrative Judge
within 5 days of the notification by the NRC staff of its grant of
access.
If challenges to the NRC staff determinations are filed, these
procedures give way to the normal process for litigating disputes
concerning access to information. The availability of interlocutory
review by the Commission of orders ruling on such NRC staff
determinations (whether granting or denying access) is governed by 10
CFR 2.311.\3\
---------------------------------------------------------------------------
\3\ Requestors should note that the filing requirements of the
NRC's E-Filing Rule (72 FR 49139; August 28, 2007) apply to appeals
of NRC staff determinations (because they must be served on a
presiding officer or the Commission, as applicable), but not to the
initial SUNSI request submitted to the NRC staff under these
procedures.
---------------------------------------------------------------------------
I. The Commission expects that the NRC staff and presiding officers
(and any other reviewing officers) will consider and resolve requests
for access to SUNSI, and motions for protective orders, in a timely
fashion in order to minimize any unnecessary delays in identifying
those petitioners who have standing and who have propounded contentions
meeting the specificity and basis requirements in 10 CFR Part 2.
Attachment 1 to this Order summarizes the general target schedule for
processing and resolving requests under these procedures.
It is so ordered.
Dated at Rockville, Maryland, this 4th day of January, 2011.
[[Page 1469]]
For the Nuclear Regulatory Commission.
Andrew L. Bates,
Acting Secretary of the Commission.
Attachment 1--General Target Schedule for Processing and Resolving
Requests for Access to Sensitive Unclassified Non-Safeguards
Information in This Proceeding
------------------------------------------------------------------------
Day Event/activity
------------------------------------------------------------------------
0........................ Publication of Federal Register notice of
hearing and opportunity to petition for
leave to intervene, including order with
instructions for access requests.
10....................... Deadline for submitting requests for access
to Sensitive Unclassified Non-Safeguards
Information (SUNSI) with information:
Supporting the standing of a potential party
identified by name and address; describing
the need for the information in order for
the potential party to participate
meaningfully in an adjudicatory proceeding.
60....................... Deadline for submitting petition for
intervention containing: (i) Demonstration
of standing; (ii) all contentions whose
formulation does not require access to SUNSI
(+25 Answers to petition for intervention;
+7 requestor/petitioner reply).
20....................... Nuclear Regulatory Commission (NRC) staff
informs the requestor of the staff's
determination whether the request for access
provides a reasonable basis to believe
standing can be established and shows need
for SUNSI. (NRC staff also informs any party
to the proceeding whose interest independent
of the proceeding would be harmed by the
release of the information.) If NRC staff
makes the finding of need for SUNSI and
likelihood of standing, NRC staff begins
document processing (preparation of
redactions or review of redacted documents).
25....................... If NRC staff finds no ``need'' or no
likelihood of standing, the deadline for
requestor/petitioner to file a motion
seeking a ruling to reverse the NRC staff's
denial of access; NRC staff files copy of
access determination with the presiding
officer (or Chief Administrative Judge or
other designated officer, as appropriate).
If NRC staff finds ``need'' for SUNSI, the
deadline for any party to the proceeding
whose interest independent of the proceeding
would be harmed by the release of the
information to file a motion seeking a
ruling to reverse the NRC staff's grant of
access.
30....................... Deadline for NRC staff reply to motions to
reverse NRC staff determination(s).
40....................... (Receipt +30) If NRC staff finds standing and
need for SUNSI, deadline for NRC staff to
complete information processing and file
motion for Protective Order and draft Non-
Disclosure Affidavit. Deadline for applicant/
licensee to file Non-Disclosure Agreement
for SUNSI.
A........................ If access granted: Issuance of presiding
officer or other designated officer decision
on motion for protective order for access to
sensitive information (including schedule
for providing access and submission of
contentions) or decision reversing a final
adverse determination by the NRC staff.
A + 3.................... Deadline for filing executed Non-Disclosure
Affidavits. Access provided to SUNSI
consistent with decision issuing the
protective order.
A + 28................... Deadline for submission of contentions whose
development depends upon access to SUNSI.
However, if more than 25 days remain between
the petitioner's receipt of (or access to)
the information and the deadline for filing
all other contentions (as established in the
notice of hearing or opportunity for
hearing), the petitioner may file its SUNSI
contentions by that later deadline.
A + 53................... (Contention receipt +25) Answers to
contentions whose development depends upon
access to SUNSI.
A + 60................... (Answer receipt +7) Petitioner/Intervenor
reply to answers.
> A + 60................. Decision on contention admission.
------------------------------------------------------------------------
[FR Doc. 2011-215 Filed 1-7-11; 8:45 am]
BILLING CODE 7590-01-P