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, 1462-1469 [2011-215]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 76, Number 6 (Monday, January 10, 2011)]
[Notices]
[Pages 1462-1469]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2011-215]


=======================================================================
-----------------------------------------------------------------------

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 https://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, https://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 
hearing.docket@nrc.gov, 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 https://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 
https://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 https://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 https://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 https://www.nrc.gov/site-help/e-submittals.html, by e-mail at 
MSHD.Resource@nrc.gov, 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 
https://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, https://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 pdr.resource@nrc.gov.

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 Hearing.Docket@nrc.gov and OGCmailcenter@nrc.gov, 
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