ML12152A316

Biweekly Sholly Report - Federal Register Notice - Fr Publication June 12, 2012

ML12152A316
Person / Time
Site:	Seabrook
Issue date:	06/12/2012
From:	Marilyn Evans Office of Nuclear Reactor Regulation
To:
Tully B
References
NRC-2010-0206
Download: ML12152A316 (11)

Text

Federal Register / Vol. 77, No. 113 / Tuesday, June 12, 2012 / Notices 35069 NUCLEAR REGULATORY I. Accessing Information and Notice of Consideration of Issuance of COMMISSION Submitting Comments Amendments to Facility Operating Licenses and Combined Licenses, A. Accessing Information

[NRC-2012-0131] Proposed No Significant Hazards Please refer to Docket ID NRC-2012- Consideration Determination, and Biweekly Notice; Applications and 0131 when contacting the NRC about Opportunity for a Hearing Amendments to Facility Operating the availability of information regarding The Commission has made a Licenses and Combined Licenses this document. You may access proposed determination that the Involving No Significant Hazards information related to this document, following amendment requests involve Considerations which the NRC possesses and is no significant hazards consideration.

publicly available, by the following Under the Commissions regulations in Background methods: Title 10 of the Code of Federal Pursuant to Section 189a.(2) of the

• Federal Rulemaking Web site: Go to Regulations (10 CFR) 50.92, this means Atomic Energy Act of 1954, as amended http://www.regulations.gov and search that operation of the facility in (the Act), the U.S. Nuclear Regulatory for Docket ID NRC-2012-0131. accordance with the proposed Commission (the Commission or NRC)

• NRCs Agencywide Documents amendment would not (1) involve a is publishing this regular biweekly Access and Management System significant increase in the probability or notice. The Act requires the (ADAMS): You may access publicly consequences of an accident previously Commission publish notice of any available documents online in the NRC evaluated; or (2) create the possibility of amendments issued, or proposed to be Library at http://www.nrc.gov/reading- a new or different kind of accident from issued and grants the Commission the rm/adams.html. To begin the search, any accident previously evaluated; or authority to issue and make select ADAMS Public Documents and (3) involve a significant reduction in a immediately effective any amendment then select Begin Web-based ADAMS margin of safety. The basis for this to an operating license or combined Search. For problems with ADAMS, proposed determination for each license, as applicable, upon a please contact the NRCs Public amendment request is shown below.

determination by the Commission that Document Room (PDR) reference staff at The Commission is seeking public such amendment involves no significant 1-800-397-4209, 301-415-4737, or by comments on this proposed hazards consideration, notwithstanding email to pdr.resource@nrc.gov. determination. Any comments received the pendency before the Commission of Documents may be viewed in ADAMS within 30 days after the date of a request for a hearing from any person. by performing a search on the document publication of this notice will be This biweekly notice includes all date and docket number. considered in making any final notices of amendments issued, or

• NRCs PDR: You may examine and determination.

proposed to be issued from May 17, Normally, the Commission will not purchase copies of public documents at 2012 to May 30, 2012. The last biweekly issue the amendment until the the NRCs PDR, Room O1-F21, One notice was published on May 29, 2012 expiration of 60 days after the date of White Flint North, 11555 Rockville (77 FR 31655). publication of this notice. The Pike, Rockville, Maryland 20852.

Commission may issue the license ADDRESSES: You may access information B. Submitting Comments amendment before expiration of the 60-and comment submissions related to day period provided that its final this document, which the NRC Please include Docket ID NRC-2012- determination is that the amendment possesses and is publicly available, by 0131 in the subject line of your involves no significant hazards searching on http://www.regulations.gov comment submission, in order to ensure consideration. In addition, the under Docket ID NRC-2012-0131. You that the NRC is able to make your Commission may issue the amendment may submit comments by the following comment submission available to the prior to the expiration of the 30-day methods: public in this docket. comment period should circumstances

• Federal Rulemaking Web site: Go to The NRC cautions you not to include change during the 30-day comment http://www.regulations.gov and search identifying or contact information in period such that failure to act in a for Docket ID NRC-2012-0131. Address comment submissions that you do not timely way would result, for example in questions about NRC dockets to Carol want to be publicly disclosed. The NRC derating or shutdown of the facility.

Gallagher; telephone: 301-492-3668; posts all comment submissions at Should the Commission take action email: Carol.Gallagher@nrc.gov. http://www.regulations.gov as well as prior to the expiration of either the

• Mail comments to: Cindy Bladey, entering the comment submissions into comment period or the notice period, it Chief, Rules, Announcements, and ADAMS, and the NRC does not edit will publish in the Federal Register a Directives Branch (RADB), Office of comment submissions to remove notice of issuance. Should the Administration, Mail Stop: TWB identifying or contact information. Commission make a final No Significant B01M, U.S. Nuclear Regulatory If you are requesting or aggregating Hazards Consideration Determination; Commission, Washington, DC 20555- comments from other persons for any hearing will take place after 0001. submission to the NRC, then you should issuance. The Commission expects that

• Fax comments to: RADB at 301- inform those persons not to include the need to take this action will occur 492-3446. identifying or contact information in very infrequently.

their comment submissions that they do Within 60 days after the date of srobinson on DSK4SPTVN1PROD with NOTICES For additional direction on accessing not want to be publicly disclosed. Your publication of this notice, any person(s) information and submitting comments, request should state that the NRC will whose interest may be affected by this see Accessing Information and not edit comment submissions to action may file a request for a hearing Submitting Comments in the remove such information before making and a petition to intervene with respect SUPPLEMENTARY INFORMATION section of the comment submissions available to to issuance of the amendment to the this document. the public or entering the comment subject facility operating license or SUPPLEMENTARY INFORMATION: submissions into ADAMS. combined license. Requests for a VerDate Mar<15>2010 22:42 Jun 11, 2012 Jkt 226001 PO 00000 Frm 00137 Fmt 4703 Sfmt 4703 E:\FR\FM\12JNN1.SGM 12JNN1

35070 Federal Register / Vol. 77, No. 113 / Tuesday, June 12, 2012 / Notices hearing and a petition for leave to sufficient information to show that a documents and access the E-Submittal intervene shall be filed in accordance genuine dispute exists with the server for any proceeding in which it is with the Commissions Rules of applicant on a material issue of law or participating; and (2) advise the Practice for Domestic Licensing fact. Contentions shall be limited to Secretary that the participant will be Proceedings in 10 CFR Part 2. matters within the scope of the submitting a request or petition for Interested person(s) should consult a amendment under consideration. The hearing (even in instances in which the current copy of 10 CFR 2.309, which is contention must be one which, if participant, or its counsel or available at the NRCs PDR, located at proven, would entitle the requestor/ representative, already holds an NRC-One White Flint North, Room O1-F21, petitioner to relief. A requestor/ issued digital ID certificate). Based upon 11555 Rockville Pike (first floor), petitioner who fails to satisfy these this information, the Secretary will Rockville, Maryland 20852. The NRC requirements with respect to at least one establish an electronic docket for the regulations are accessible electronically contention will not be permitted to hearing in this proceeding if the from the NRC Library on the NRCs Web participate as a party. Secretary has not already established an site at http://www.nrc.gov/reading-rm/ Those permitted to intervene become electronic docket.

doc-collections/cfr/. If a request for a parties to the proceeding, subject to any Information about applying for a hearing or petition for leave to intervene limitations in the order granting leave to digital ID certificate is available on the is filed by the above date, the intervene, and have the opportunity to NRCs public Web site at http://

Commission or a presiding officer participate fully in the conduct of the www.nrc.gov/site-help/e-submittals/

designated by the Commission or by the hearing. apply-certificates.html. System Chief Administrative Judge of the If a hearing is requested, the requirements for accessing the E-Atomic Safety and Licensing Board Commission will make a final Submittal server are detailed in the Panel, will rule on the request and/or determination on the issue of no NRCs Guidance for Electronic petition; and the Secretary or the Chief significant hazards consideration. The Submission, which is available on the Administrative Judge of the Atomic final determination will serve to decide agencys public Web site at http://

Safety and Licensing Board will issue a when the hearing is held. If the final www.nrc.gov/site-help/e-notice of a hearing or an appropriate determination is that the amendment submittals.html. Participants may order. request involves no significant hazards attempt to use other software not listed As required by 10 CFR 2.309, a consideration, the Commission may on the Web site, but should note that the petition for leave to intervene shall set issue the amendment and make it NRCs E-Filing system does not support forth with particularity the interest of immediately effective, notwithstanding unlisted software, and the NRC Meta the petitioner in the proceeding, and the request for a hearing. Any hearing System Help Desk will not be able to how that interest may be affected by the held would take place after issuance of offer assistance in using unlisted results of the proceeding. The petition the amendment. If the final software.

should specifically explain the reasons determination is that the amendment If a participant is electronically why intervention should be permitted request involves a significant hazards submitting a document to the NRC in with particular reference to the consideration, then any hearing held accordance with the E-Filing rule, the following general requirements: (1) The would take place before the issuance of participant must file the document name, address, and telephone number of any amendment. using the NRCs online, Web-based the requestor or petitioner; (2) the All documents filed in NRC submission form. In order to serve nature of the requestors/petitioners adjudicatory proceedings, including a documents through Electronic right under the Act to be made a party request for hearing, a petition for leave Information Exchange System, users to the proceeding; (3) the nature and to intervene, any motion or other will be required to install a Web extent of the requestors/petitioners document filed in the proceeding prior browser plug-in from the NRCs Web property, financial, or other interest in to the submission of a request for site. Further information on the Web-the proceeding; and (4) the possible hearing or petition to intervene, and based submission form, including the effect of any decision or order which documents filed by interested installation of the Web browser plug-in, may be entered in the proceeding on the governmental entities participating is available on the NRCs public Web requestors/petitioners interest. The under 10 CFR 2.315(c), must be filed in site at http://www.nrc.gov/site-help/e-petition must also identify the specific accordance with the NRC E-Filing rule submittals.html.

contentions which the requestor/ (72 FR 49139; August 28, 2007). The E- Once a participant has obtained a petitioner seeks to have litigated at the Filing process requires participants to digital ID certificate and a docket has proceeding. submit and serve all adjudicatory been created, the participant can then Each contention must consist of a documents over the internet, or in some submit a request for hearing or petition specific statement of the issue of law or cases to mail copies on electronic for leave to intervene. Submissions fact to be raised or controverted. In storage media. Participants may not should be in Portable Document Format addition, the requestor/petitioner shall submit paper copies of their filings (PDF) in accordance with NRC guidance provide a brief explanation of the bases unless they seek an exemption in available on the NRC public Web site at for the contention and a concise accordance with the procedures http://www.nrc.gov/site-help/e-statement of the alleged facts or expert described below. submittals.html. A filing is considered opinion which support the contention To comply with the procedural complete at the time the documents are and on which the requestor/petitioner requirements of E-Filing, at least 10 submitted through the NRCs E-Filing intends to rely in proving the contention days prior to the filing deadline, the system. To be timely, an electronic srobinson on DSK4SPTVN1PROD with NOTICES at the hearing. The requestor/petitioner participant should contact the Office of filing must be submitted to the E-Filing must also provide references to those the Secretary by email at system no later than 11:59 p.m. Eastern specific sources and documents of hearing.docket@nrc.gov, or by telephone Time on the due date. Upon receipt of which the petitioner is aware and on at 301-415-1677, to request (1) a digital a transmission, the E-Filing system which the requestor/petitioner intends identification (ID) certificate, which time-stamps the document and sends to rely to establish those facts or expert allows the participant (or its counsel or the submitter an email notice opinion. The petition must include representative) to digitally sign confirming receipt of the document. The VerDate Mar<15>2010 22:42 Jun 11, 2012 Jkt 226001 PO 00000 Frm 00138 Fmt 4703 Sfmt 4703 E:\FR\FM\12JNN1.SGM 12JNN1

Federal Register / Vol. 77, No. 113 / Tuesday, June 12, 2012 / Notices 35071 E-Filing system also distributes an email available to the public at http:// No. ML062860320). The changes are notice that provides access to the ehd1.nrc.gov/ehd/, unless excluded consistent with NUREG-1432, document to the NRCs Office of the pursuant to an order of the Commission, Standard Technical Specifications General Counsel and any others who or the presiding officer. Participants are Combustion Engineering Plants, have advised the Office of the Secretary requested not to include personal Revision 4 (Agencywide Documents that they wish to participate in the privacy information, such as social Access and Management System proceeding, so that the filer need not security numbers, home addresses, or (ADAMS) Accession No.

serve the documents on those home phone numbers in their filings, ML12102A165).

participants separately. Therefore, unless an NRC regulation or other law Basis for proposed no significant applicants and other participants (or requires submission of such hazards consideration determination:

their counsel or representative) must information. With respect to As required by 10 CFR 50.91(a), the apply for and receive a digital ID copyrighted works, except for limited licensee has provided its analysis of the certificate before a hearing request/ excerpts that serve the purpose of the issue of no significant hazards petition to intervene is filed so that they adjudicatory filings and would consideration, which is presented can obtain access to the document via constitute a Fair Use application, below:

the E-Filing system. participants are requested not to include 1. Does the proposed amendment involve A person filing electronically using copyrighted materials in their a significant increase in the probability or the agencys adjudicatory E-Filing submission. consequences of an accident previously system may seek assistance by Petitions for leave to intervene must evaluated?

contacting the NRC Meta System Help be filed no later than 60 days from the Response: No.

Desk through the Contact Us link date of publication of this notice. Non- The proposed change eliminates the use of located on the NRC Web site at http:// timely filings will not be entertained the defined term CORE ALTERATIONS from www.nrc.gov/site-help/e- absent a determination by the presiding the Technical Specifications. CORE submittals.html, by email at officer that the petition or request ALTERATIONS are not an initiator of any accident previously evaluated except a fuel MSHD.Resource@nrc.gov, or by a toll- should be granted or the contentions handling accident. The revised Technical free call at 1-866 672-7640. The NRC should be admitted, based on a Specifications that protect the initial Meta System Help Desk is available balancing of the factors specified in 10 conditions of a fuel handling accident also between 8 a.m. and 8 p.m., Eastern CFR 2.309(c)(1)(i)-(viii). require the suspension of movement of Time, Monday through Friday, For further details with respect to this irradiated fuel assemblies. Suspending excluding government holidays. license amendment application, see the movement of irradiated fuel assemblies Participants who believe that they application for amendment which is protects the initial condition of a fuel have a good cause for not submitting available for public inspection at the handling accident and, therefore, suspension documents electronically must file an NRCs PDR, located at One White Flint of CORE ALTERATIONS is not required.

Suspension of CORE ALTERATIONS does exemption request, in accordance with North, Room O1-F21, 11555 Rockville not provide mitigation of any accident 10 CFR 2.302(g), with their initial paper Pike (first floor), Rockville, Maryland previously evaluated. Therefore, CORE filing requesting authorization to 20852. Publicly available documents ALTERATIONS do not affect the initiators of continue to submit documents in paper created or received at the NRC are the accidents previously evaluated and format. Such filings must be submitted accessible electronically through suspension of CORE ALTERATIONS does by: (1) First class mail addressed to the ADAMS in the NRC Library at http:// not affect the mitigation of the accidents Office of the Secretary of the www.nrc.gov/reading-rm/adams.html. previously evaluated.

Commission, U.S. Nuclear Regulatory Persons who do not have access to Therefore, the proposed change does not Commission, Washington, DC 20555- ADAMS or who encounter problems in involve a significant increase in the 0001, Attention: Rulemaking and accessing the documents located in probability or consequences of an accident previously evaluated.

Adjudications Staff; or (2) courier, ADAMS, should contact the NRCs PDR 2. Does the proposed change create the express mail, or expedited delivery Reference staff at 1-800-397-4209, 301- possibility of a new or different kind of service to the Office of the Secretary, 415-4737, or by email to accident from any accident previously Sixteenth Floor, One White Flint North, pdr.resource@nrc.gov. evaluated?

11555 Rockville Pike, Rockville, Response: No.

Maryland, 20852, Attention: Arizona Public Service Company, et al., No new or different accidents result from Rulemaking and Adjudications Staff. Docket Nos. STN 50-528, STN 50-529, utilizing the proposed change. The changes Participants filing a document in this and STN 50-530, Palo Verde Nuclear do not involve a physical modification of the manner are responsible for serving the Generating Station, Units 1, 2, and 3, plant (i.e., no new or different type of document on all other participants. Maricopa County, Arizona equipment will be installed) or a significant Date of amendment request: March 8, change in the methods governing normal Filing is considered complete by first-plant operation. In addition, the changes do class mail as of the time of deposit in 2012. not impose any new or different the mail, or by courier, express mail, or Description of amendment request: requirements. The changes do not alter expedited delivery service upon The amendments would eliminate the assumptions made in the safety analysis. The depositing the document with the use of the term CORE ALTERATIONS proposed changes are consistent with the provider of the service. A presiding throughout the Technical Specifications safety analysis assumptions.

officer, having granted an exemption (TSs). The proposed amendment Therefore, the proposed change does not request from using E-Filing, may require incorporates changes reflected in create the possibility of a new or different a participant or party to use E-Filing if Technical Specification Task Force kind of accident from any previously srobinson on DSK4SPTVN1PROD with NOTICES the presiding officer subsequently (TSTF) Change Traveler TSTF-471-A, evaluated.

3. Does the proposed amendment involve determines that the reason for granting Revision 1, Eliminate use of term a significant reduction in a margin of safety?

the exemption from use of E-Filing no CORE ALTERATIONS in ACTIONS and Response: No.

longer exists. Notes. The U.S. Nuclear Regulatory Only two accidents are postulated to occur Documents submitted in adjudicatory Commission (NRC) staff reviewed and during plant conditions where CORE proceedings will appear in the NRCs approved TSTF-471 by letter dated ALTERATIONS may be made: a fuel electronic hearing docket which is December 7, 2006 (ADAMS Accession handling accident and a boron dilution VerDate Mar<15>2010 22:42 Jun 11, 2012 Jkt 226001 PO 00000 Frm 00139 Fmt 4703 Sfmt 4703 E:\FR\FM\12JNN1.SGM 12JNN1

35072 Federal Register / Vol. 77, No. 113 / Tuesday, June 12, 2012 / Notices accident. Suspending movement of irradiated Bases for Diesel Fuel Oil, Revision 0. To Criterion 3 fuel assemblies prevents a fuel handling adopt changes consistent with the content of Does the proposed amendment involve a accident. Also requiring the suspension of TSTF-374 for use in the custom TS of MPS2, significant reduction in the margin of safety?

CORE ALTERATIONS is a redundant the existing MPS2 diesel fuel oil testing Response: No.

requirement to suspending movement of program will be modified. These changes The proposed changes are consistent with irradiated fuel assemblies and does not replace the criteria of Water and sediment the content of TSTF-374 for use in the increase the margin of safety. CORE < 0.05% with the criteria of A clear and custom TS of MPS2. These changes remove ALTERATIONS have no effect on a boron bright appearance with proper color or a specific ASTM standard references and a dilution accident. Core components are not water and sediment content within limits preventive maintenance cleaning involved in the initiation or mitigation of a and remove specific American Society for requirement from TS since the references and boron dilution accident and the SHUTDOWN Testing and Materials (ASTM) standard requirements are already specified in MARGIN limit is based on assuming the references from TS. licensee-controlled documents. The proposed worse-case configuration of the core The proposed changes do not adversely changes provide the flexibility needed to components. affect accident initiators or precursors nor improve fuel oil sampling and analysis Therefore, CORE ALTERATIONS have no alter the design assumptions, conditions, and methodologies while maintaining sufficient effect on the margin of safety related to a configuration of the facility or the manner in controls to ensure continued quality of the boron dilution accident. which the plant is operated and maintained. fuel oil. The margin of safety provided to the The NRC staff has reviewed the The proposed changes do not adversely affect DGs by these detailed fuel specifications is licensees analysis and, based on that the ability of structures, systems, and unaffected by the proposed changes since components (SSCs) to perform their intended there continue to be TS requirements to review, it appears that the three safety function to mitigate the consequences ensure fuel oil is of the appropriate quality standards of 10 CFR 50.92(c) are for emergency DG use and DG operability is of an initiating event within the assumed satisfied. Therefore, the NRC staff acceptance limits. The proposed changes do unaffected.

proposes to determine that the request not affect the source term, containment for amendments involves no significant The NRC staff has reviewed the isolation, or radiological release assumptions licensees analysis and, based on this hazards consideration. used in evaluating the radiological Attorney for licensee: Michael G. review, it appears that the three consequences of any accident previously Green, Senior Regulatory Counsel, evaluated. Further, the proposed changes do standards of 10 CFR 50.92(c) are Pinnacle West Capital Corporation, P.O. not increase the types and amounts of satisfied. Therefore, the NRC staff Box 52034, Mail Station 8695, Phoenix, radioactive effluent that may be released proposes to determine that the Arizona 85072-2034. offsite, nor significantly increase individual amendment request involves no NRC Branch Chief: Michael T. or cumulative occupational/public radiation significant hazards consideration.

Markley. exposures. Attorney for licensee: Lillian M.

Therefore, the changes do not involve a Cuoco, Senior Counsel, Dominion Dominion Nuclear Connecticut, Inc., significant increase in the probability or Resources Services, Inc., 120 Tredegar Docket No. 50-336, Millstone Power consequences or any accident previously Street, RS-2, Richmond, VA 23219.

Station, Unit 2, New London County, evaluated. NRC Branch Chief: George Wilson.

Connecticut Criterion 2 Exelon Generation Company, LLC Date of amendment request: April 13, Does the proposed amendment create the (EGC), Docket Nos. STN 50-456 and 2012. possibility of a new or different kind of accident from any accident previously STN 50-457, Braidwood Station, Units 1 Description of amendment request: and 2 (Braidwood), Will County, Illinois, evaluated?

The proposed amendment would revise Response: No. Docket Nos. STN 50-454 and STN 50-the Millstone Power Station, Unit 2 The proposed changes are used to provide 455, Byron Station, Units 1 and 2 (MPS2) Technical Specification (TS) operational flexibility regarding evolving (Byron), Ogle County, Illinois requirements related to diesel fuel oil industry standards while maintaining Date of amendment request: March testing consistent with NUREG-1432, operational conditions which are consistent 20, 2012.

Rev. 3.1, Standard Technical with the design basis. Removing of specific details from TS, since the details are already Description of amendment request:

Specifications, Combustion Engineering specified in licensee-controlled documents, The proposed amendment would Plants, December 1, 1995, and NRC provides the flexibility needed to maintain modify Braidwood and Byron Technical approved Technical Specification Task state-of-the-art technology in fuel oil Specifications to permanently exclude Force (TSTF) TSTF-374, Revision to sampling and analysis methodology. The portions of the steam generator (SG)

TS 5.5.13 and Associated TS Bases for procedural details associated with the tube below the top of the SG tubesheet Diesel Fuel Oil, Revision 0. involved specifications that are removed from periodic SG tube inspections and Basis for proposed no significant from TS and residing in licensee-controlled plugging or repair for Braidwood, Unit hazards consideration determination: documents are not required to be in the TS 2 and for Byron, Unit 2. In addition, the As required by Title 10 of the Code of to provide adequate protection of the public proposed amendment would revise TS Federal Regulations (10 CFR) 50.91(a), health and safety, since the TS still retains the requirement for compliance with 5.6.9 to remove reference to the the licensee has provided its analysis of applicable standards. The changes do not previous temporary alternate repair the issue of no significant hazards involve a physical alteration of the plant (i.e., criteria and provide reporting consideration, which is presented no new or different type of equipment will requirements specific to the permanent below:

be installed) or a change in the methods alternate repair criteria.

Criterion 1 governing normal plant operation in the Basis for proposed no significant Does the proposed amendment involve a provision, maintaining, or use of diesel fuel hazards consideration determination:

srobinson on DSK4SPTVN1PROD with NOTICES significant increase in the probability or oil. The requirements retained in the TS As required by 10 CFR 50.91(a), the consequences of an accident previously continue to require testing of the diesel fuel licensee has provided its analysis of the evaluated? oil to ensure the proper functioning of the issue of no significant hazards Response: No. DGs.

consideration, which is presented The proposed changes modify the TS Therefore, the changes do not create the requirements related to diesel fuel oil testing possibility of a new or different kind of below:

consistent with NRC approved TSTF-374, accident from any accident previously 1. Does the proposed change involve a Revision to TS 5.5.13 and Associated TS evaluated. significant increase in the probability or VerDate Mar<15>2010 22:42 Jun 11, 2012 Jkt 226001 PO 00000 Frm 00140 Fmt 4703 Sfmt 4703 E:\FR\FM\12JNN1.SGM 12JNN1

Federal Register / Vol. 77, No. 113 / Tuesday, June 12, 2012 / Notices 35073 consequences of an accident previously Primary-to-secondary leakage from tube For the CM assessment, the component of evaluated? degradation in the tubesheet area during leakage from the prior cycle from below the Response: No. operating and accident conditions is H* distance will be multiplied by a factor of The previously analyzed accidents are restricted due to contact of the tube with the 3.11 and added to the total leakage from any initiated by the failure of plant structures, tubesheet. The leakage is modeled as flow other source and compared to the allowable systems, or components. The proposed through a porous medium through the use of accident induced leakage limit. For the OA, change that alters the steam generator (SG) the Darcy equation. The leakage model is the difference in the leakage between the inspection and reporting criteria does not used to develop a relationship between allowable leakage and the accident induced have a detrimental impact on the integrity of operational leakage and leakage at accident leakage from sources other than the tubesheet any plant structure, system, or component conditions that is based on differential expansion region will be divided by 3.11 and that initiates an analyzed event. The pressure across the tubesheet and the compared to the observed operational proposed change will not alter the operation viscosity of the fluid. A leak rate ratio was leakage.

of, or otherwise increase the failure developed to relate the leakage at operating Based on the above, the performance probability of any plant equipment that conditions to leakage at accident conditions. criteria of NEI-97-06, Revision 3, and draft initiates an analyzed accident. Since the fluid viscosity is based on fluid RG 1.121 continue to be met and the Of the various accidents previously temperature and it is shown that for the most proposed change does not involve a evaluated, the proposed changes only affect limiting accident, the fluid temperature does significant increase in the probability or the steam generator tube rupture (SGTR), not exceed the normal operating temperature consequences of the applicable accidents postulated steam line break (SLB), feedwater and therefore the viscosity ratio is assumed previously evaluated.

line break (FLB), locked rotor and control rod to be 1.0. Therefore, the leak rate ratio is a 2. Does the proposed change create the ejection accident evaluations. Loss-of-coolant function of the ratio of the accident possibility of a new or different kind of accident (LOCA) conditions cause a differential pressure and the normal accident from any accident previously compressive axial load to act on the tube. operating differential pressure. evaluated?

Therefore, since the LOCA tends to force the The leakage factor of 1.93 for Braidwood Response: No.

tube into the tubesheet rather than pull it out, Station Unit 2 and Byron Station Unit 2, for The proposed change does not introduce it is not a factor in this amendment request. a postulated SLB/FLB, has been calculated as any changes or mechanisms that create the Another faulted load consideration is a safe shown in Table 9-7 of WCAP-17072-P, possibility of a new or different kind of shutdown earthquake (SSE); however, the Revision 0. However, EGC Braidwood Station accident. Tube bundle integrity is expected seismic analysis of Model D5 SGs has shown Unit 2 and Byron Station Unit 2 will apply to be maintained for all plant conditions that axial loading of the tubes is negligible a factor of 3.11 as determined by upon implementation of the permanent during an SSE. Westinghouse evaluation LTR-SGMP alternate repair criteria. The proposed change During the SGTR event, the required 100 P-Attachment, Revision 1, to the normal does not introduce any new equipment or structural integrity margins of the SG tubes operating leakage associated with the any change to existing equipment. No new and the tube-to-tubesheet joint over the H* tubesheet expansion region in the condition effects on existing equipment are created nor distance will be maintained. Tube rupture in monitoring (CM) and operational assessment are any new malfunctions introduced.

tubes with cracks within the tubesheet is (OA). The leakage factor of 3.11 applies Therefore, based on the above evaluation, precluded by the constraint provided by the specifically to Byron Unit 2 and Braidwood the proposed changes do not create the presence of the tubesheet and the tube-to- Unit 2, both hot and cold legs, in Table possibility of a new or different kind of tubesheet joint. Tube burst cannot occur RAI24-2 of LTR-SGMP-09-100 P- accident from any accident previously within the thickness of the tubesheet. The Attachment, Revision 1. Through application evaluated.

tube-to-tubesheet joint constraint results from of the limited tubesheet inspection scope, the 3. Does the proposed change involve a the hydraulic expansion process, thermal existing operating leakage limit provides significant reduction in a margin of safety?

expansion mismatch between the tube and assurance that excessive leakage (i.e., greater Response: No.

tubesheet, and from the differential pressure than accident analysis assumptions) will not The proposed change defines the safety between the primary and secondary side, and occur. The assumed accident induced leak significant portion of the SG tube that must tubesheet rotation. Based on this design, the rate limit is 0.5 gallons per minute at room be inspected and repaired. WCAP-17072-P, structural margins against burst, as discussed temperature (gpmRT) for the faulted SG and Revision 0, as modified by WCAP-17330-P, in draft Regulatory Guide (RG) 1.121, Bases 0.218 gpmRT for each of the unfaulted SGs Revision 1, identifies the specific inspection for Plugging Degraded PWR Steam Generator for accidents that assume a faulted SG. These depth below which any type tube Tubes, and TS 5.5.9, are maintained for both accidents are the SLB and the locked rotor degradation has no impact on the normal and postulated accident conditions. with a stuck open PORV. The assumed performance criteria in NEI 97-06, Revision The proposed change has no impact on the accident induced leak rate limit for accidents 3, Steam Generator Program Guidelines.

structural or leakage integrity of the portion that do not assume a faulted SG is 1.0 gpmRT The proposed change that alters the SG of the tube outside of the tubesheet. The for all SGs. These accidents are the locked inspection and reporting criteria maintains proposed change maintains structural and rotor and control rod ejection. the required structural margins of the SG leakage integrity of the SG tubes consistent No leakage factor will be applied to the tubes for both normal and accident with the performance criteria of TS 5.5.9. locked rotor or control rod ejection transients conditions. NEI 97-06, and draft RG 1.121 Therefore, the proposed change results in no due to their short duration, since the are used as the bases in the development of significant increase in the probability of the calculated leak rate ratio is less than 1.0. the limited tubesheet inspection depth occurrence of a SGTR accident. The TS 3.4.13 operational leak rate limit is methodology for determining that SG tube At normal operating pressures, leakage 150 gallons per day (gpd) (0.104 gpmRT) integrity considerations are maintained from tube degradation below the proposed through any one SG. Consequently, there is within acceptable limits. Draft RG 1.121 limited inspection depth is limited by the sufficient margin between accident leakage describes a method acceptable to the NRC for tube-to-tubesheet crevice. Consequently, and allowable operational leakage. The meeting General Design Criteria (GDC) 14, negligible normal operating leakage is maximum accident leak rate ratio for the Reactor Coolant Pressure Boundary, GDC expected from degradation below the Model D5 design SGs is 1.93 as indicated in 15, Reactor Coolant System Design, GDC inspected depth within the tubesheet region. WCAP-17072-P, Revision 0, Table 9-7. 31, Fracture Prevention of Reactor Coolant The consequences of an SGTR event are However, EGC will use the more conservative Pressure Boundary, and GDC 32, srobinson on DSK4SPTVN1PROD with NOTICES not affected by the primary-to-secondary value of 3.11 accident leak rate ratio for the Inspection of Reactor Coolant Pressure leakage flow during the event as primary-to- most limiting SG model design identified in Boundary, by reducing the probability and secondary leakage flow through a postulated Table RAI24-2 of LTRSGMP-09-100 P- consequences of a SGTR. Draft RG 1.121 tube that has been pulled out of the tubesheet Attachment Revision 1. This results in concludes that by determining the limiting is essentially equivalent to a severed tube. significant margin between the safe conditions for tube wall degradation, the Therefore, the proposed change does not conservatively estimated accident leakage probability and consequences of a SGTR are result in a significant increase in the and the allowable accident leakage (0.5 reduced. This draft RG uses safety factors on consequences of a SGTR. gpmRT). loads for tube burst that are consistent with VerDate Mar<15>2010 22:42 Jun 11, 2012 Jkt 226001 PO 00000 Frm 00141 Fmt 4703 Sfmt 4703 E:\FR\FM\12JNN1.SGM 12JNN1

35074 Federal Register / Vol. 77, No. 113 / Tuesday, June 12, 2012 / Notices the requirements of Section III of the previously been established on a one- tube-to-tubesheet crevice. Consequently, American Society of Mechanical Engineers cycle basis. negligible normal operating leakage is (ASME) Code. The proposed amendment constitutes expected from degradation below the For axially oriented cracking located a redefinition of the SG tube primary-to- inspected depth within the tubesheet region.

within the tubesheet, tube burst is precluded The consequences of an SGTR event are not secondary pressure boundary and affected by the primary-to-secondary leakage due to the presence of the tubesheet. For circumferentially oriented cracking, WCAP- defines the safety significant portion of flow during the event as primary-to-17072-P, Revision 0, as modified by WCAP- the tube that must be inspected or secondary leakage flow through a postulated 17330-P, Revision 1, defines a length of plugged. Tube flaws detected below the tube that has been pulled out of the tubesheet degradation-free expanded tubing that safety significant portion of the tube are is essentially equivalent to a severed tube.

provides the necessary resistance to tube not required to be plugged. Allowing Therefore, the proposed change does not pullout due to the pressure induced forces, flaws in the non-safety significant result in a significant increase in the with applicable safety factors applied. consequences of [an] SGTR.

portion of the tube to remain in service The probability of [an] SLB is unaffected Application of the limited hot and cold leg minimizes unnecessary tube plugging tubesheet inspection criteria will preclude by the potential failure of a steam generator and maintains the safety margin of the tube as the failure of tube is not an initiator unacceptable primary-to-secondary leakage during all plant conditions. The methodology steam generators to perform the safety for [an] SLB event.

for determining leakage as described in function to maintain the reactor coolant The leakage factor of 3.16 for CPNPP Unit WCAP-17072-P, Revision 0, as modified by pressure boundary, maintain reactor 2, for a postulated SLB/FLB, has been LTR-SGMP-09-100 P-Attachment\ shows coolant flow, and maintain primary to calculated as described in Westinghouse

[Electric Company, LLC] Letter LTR-SGMP-that significant margin exists between an secondary heat transfer.09-100 [N]PAttachment, Response to acceptable level of leakage during normal Basis for proposed no significant NRC Request for Additional Information on operating conditions that ensures meeting the hazards consideration determination: H*; Model F and Model D5 Steam SLB accident-induced leakage assumption As required by 10 CFR 50.91(a), the Generators, dated August 12, 2009 and the TS leakage limit of 150 gpd. licensee has provided its analysis of the [(ADAMS Accession No. ML101730391)],

Based on the above, it is concluded that the issue of no significant hazards and is shown in Revised Table 9-7 of this proposed changes do not result in any same document. Specifically, for the reduction in a margin of safety. consideration, which is presented below: condition monitoring (CM) assessment, the Based on the above, EGC concludes that component of leakage from the prior cycle the proposed change presents no significant 1. Do the proposed changes involve a from below the H* distance will be hazards consideration under the standards significant increase in the probability or multiplied by a factor of 3.16 and added to set forth in 10 CFR 50.92(c), and, consequences of an accident previously the total leakage from any other source and accordingly, a finding of no significant evaluated? compared to the allowable accident induced hazards consideration is justified. Response: No. leakage limit. For the operational assessment Of the accidents previously evaluated, the (OA), the difference in the leakage between The NRC staff has reviewed the limiting transients with consideration to the the allowable leakage and the accident licensees analysis and, based on this proposed change to the SG tube inspection induced leakage from sources other than the review, it appears that the three and repair criteria are the steam generator tubesheet expansion region will be divided standards of 10 CFR 50.92(c) are tube rupture (SGTR) event, the steam line by 3.16 and compared to the observed satisfied. Therefore, the NRC staff break (SLB), and the feed line break (FLB) operational leakage. The accident-induced proposes to determine that the postulated accidents. leak rate limit for CPNPP Unit 2 is 1.0 gpm requested amendments involve no The required structural integrity margins of [gallons per minute]. The TS operational leak significant hazards consideration. the SG tubes and the tube-to-tubesheet joint rate limit through any one steam generator is over the H* distance will be maintained. 150 gpd [gallons per day] (0.1 gpm).

Attorney for licensee: Mr. Bradley J. Tube rupture in tubes with cracks within the Consequently, there is significant margin Fewell, Associate General Counsel, tubesheet is precluded by the constraint between accident leakage and allowable Exelon Nuclear, 4300 Winfield Road provided by the presence of the tubesheet operational leakage. The SLB/FLB overall Warrenville, IL 60555. and the tube-to-tubesheet joint. Tube burst leakage factor is 3.16 resulting in significant NRC Branch Chief: Jacob I. cannot occur within the thickness of the margin between the conservatively estimated Zimmerman. tubesheet. The tube-to-tubesheet joint accident induced leakage and the allowable constraint results from the hydraulic accident leakage.

Luminant Generation Company LLC, expansion process, thermal expansion No leakage factor was applied to the locked Docket Nos. 50-445 and 50-446, mismatch between the tube and tubesheet, rotor or control rod ejection transients due to Comanche Peak Nuclear Power Plant, differential pressure between the primary their short duration.

Units 1 and 2, Somervell County, Texas and secondary side, and tubesheet rotation. The previously analyzed accidents are Based on this design, the structural margins initiated by the failure of plant structures, Date of amendment request: March against burst, as discussed in Regulatory systems, or components. The proposed 28, 2012. Guide (RG) 1.121, Bases for Plugging change that alters the SG inspection and Brief description of amendment: The Degraded PWR [Pressurized Water Reactor] reporting criteria does not have a detrimental amendment would revise Technical Steam Generator Tubes, [(Agencywide impact on the integrity of any plant structure, Documents Access and Management System system, or component that initiates an Specification (TS) 5.5.9, Unit 1 Model (ADAMS) Accession No. ML082120667)] and analyzed event. The proposed change will D76 and Unit 2 Model D5 Steam TS 5.5.9 are maintained for both normal and not alter the operation of, or otherwise Generator (SG) Program, to postulated accident conditions. increase the failure probability of any plant permanently exclude portions of the The proposed change has no impact on the equipment that initiates an analyzed Comanche Peak Nuclear Power Plant structural or leakage integrity of the portion accident.

(CPNPP), Unit 2, Model D5 SG tubes of the tube outside of the tubesheet. The Therefore, the proposed changes do not below the top of the SG tubesheet from proposed change maintains structural and involve a significant increase in the srobinson on DSK4SPTVN1PROD with NOTICES periodic SG tube inspections. In leakage integrity of the SG tubes consistent probability or consequences of an accident with the performance criteria in TS 5.5.9. previously evaluated.

addition, this amendment would revise Therefore, the proposed change results in no 2. Do the proposed changes create the TS 5.6.9, Unit 1 Model D76 and Unit significant increase in the probability of the possibility of a new or different kind of 2 Model D5 Steam Generator Tube occurrence of [an] SGTR accident. accident from any previously evaluated?

Inspection Report, to provide At normal operating pressures, leakage Response: No.

permanent reporting requirements from tube degradation below the proposed The proposed change that alters the steam specific to CPNPP, Unit 2, that have limited inspection depth is limited by the generator inspection and reporting criteria VerDate Mar<15>2010 22:42 Jun 11, 2012 Jkt 226001 PO 00000 Frm 00142 Fmt 4703 Sfmt 4703 E:\FR\FM\12JNN1.SGM 12JNN1

Federal Register / Vol. 77, No. 113 / Tuesday, June 12, 2012 / Notices 35075 does not introduce any new equipment, Attorney for licensee: Timothy P. of depressurizing the reactor pressure vessel create new failure modes for existing Matthews, Esq., Morgan, Lewis and (RPV). This will protect the reactor vessel equipment, or create any new limiting single Bockius, 1800 M Street NW., from overpressurization and allow the failures. Plant operation will not be altered, combination of the Low Pressure Coolant Washington, DC 20036. Injection and Core Spray Systems to inject and all safety functions will continue to perform as previously assumed in accident NRC Branch Chief: Michael T. into the RPV as designed. The LLS relief analyses. Markley. logic causes two LLS relief valves to be Therefore, the proposed change does not opened at a lower pressure than the relief NextEra Energy Duane Arnold, LLC, mode pressure setpoints and causes the LLS create the possibility of a new or different kind of accident from any previously Docket No. 50-331, Duane Arnold relief valves to stay open longer, such that evaluated. Energy Center (DAEC), Linn County, reopening of more than one valve is

3. Do the proposed changes involve a Iowa prevented on subsequent actuations. Thus, significant reduction in the margin of safety? the LLS relief function prevents excessive Date of amendment request: short duration SRV cycling, which limits Response: No.

September 29, 2011, as supplemented induced thrust loads on the SRV discharge The proposed change that alters the steam generator inspection and reporting criteria by letter dated March 12, 2012. line for subsequent actuations of the relief maintains the required structural margins of Description of amendment request: valve. The proposed changes do not affect the SG tubes for both normal and accident The proposed amendment would revise any function related to the safety mode of the conditions. Nuclear Energy Institute [(NEI) the DAEC Technical Specifications dual function SRVs. The proposed changes document NEI] 97-06, Rev. 3, Steam (TSs) by modifying existing involve the manner in which the subject Generator Program Guidelines, and NRC Surveillance Requirements (SRs) valves are tested, and have no effect on the Regulatory Guide (RG) 1.121, Bases for types or amounts of radiation released or the regarding various modes of operation of predicted offsite doses in the event of an Plugging Degraded PWR Steam Generator the main steam safety/relief valves accident. The proposed testing requirements Tubes, are used as the bases in the development of the limited tubesheet (SRVs). The proposed amendment are sufficient to provide confidence that inspection depth methodology for would modify the TS requirements for these valves are capable of performing their determining that SG tube integrity testing of the SRVs by replacing the intended safety functions.

considerations are maintained within current requirement to manually actuate In addition, an inadvertent opening of an acceptable limits. RG 1.121 describes a each SRV during plant startup with a SRV is an analyzed event in the DAEC method acceptable to the NRC for meeting UFSAR [Updated Final Safety Analysis series of overlapping tests that Report] (Section 15.1.7.2), as well as the General Design Criteria (GDC) 14, Reactor demonstrate the required functions of coolant pressure boundary, GDC 15, assumption of a single SRV failure to open successive valve stages. Elimination of on demand in other transients and accidents, Reactor coolant system design, GDC 31, Fracture prevention of reactor coolant the manual actuation requirement at as appropriate (e.g., one ADS valve failure in pressure boundary, and GDC 32, low reactor pressure and steam flow the LOCA [loss-of-coolant accident] analysis).

Inspection of reactor coolant pressure decreases the potential for SRV leakage Since the proposed testing requirements do boundary, by reducing the probability and and spurious SRV opening. not alter the assumptions for any analyzed consequences of a SGTR. RG 1.121 concludes transient or accident, the radiological Basis for proposed no significant that by determining the limiting safe consequences of any accident previously hazards consideration determination: evaluated are not increased.

conditions for tube wall degradation, the As required by 10 CFR 50.91(a), the probability and consequences of a SGTR are Therefore, the change does not involve a licensee has provided its analysis of the significant increase in the probability or reduced. RG 1.121 uses safety factors on loads for tube burst that are consistent with issue of no significant hazards consequences of an accident previously the requirements of Section III of the consideration, which is presented evaluated.

American Society of Mechanical Engineers below: 2. Does the proposed amendment create (ASME) Code. the possibility of a new or different kind of

1. Does the proposed amendment involve accident from any previously evaluated?

For axially oriented cracking located a significant increase in the probability or within the tubesheet, tube burst is precluded Response: No.

consequences of any accident previously The proposed changes do not affect the due to the presence of the tubesheet. For evaluated?

circumferentially oriented cracking, the H* assumed accident performance of the main Response: No. steam SRVs, nor any plant structure, system, Analysis documented in Section 4.1 [of the The proposed changes modify TS SR or component previously evaluated. The application dated March 28, 2012] defines a 3.4.3.2, SR 3.5.1.9, and SR 3.6.1.5.1 to proposed changes do not install any new length of degradation-free expanded tubing provide an alternative means for testing the equipment, and installed equipment is not that provides the necessary resistance to tube main steam SRVs, ADS [Automatic being operated in a new or different manner.

pullout due to the pressure induced forces, Depressurization System] valves, and LLS The proposed change in test methodology with applicable safety factors applied. [Low-Low Set] relief valves. Accidents are will ensure that the valves remain capable of Application of the limited hot and cold leg initiated by the malfunction of plant performing their safety functions due to tubesheet inspection criteria will preclude equipment, or the catastrophic failure of meeting the testing requirements of the unacceptable primary-to-secondary leakage plant structures, systems, or components. American Society of Mechanical Engineers during all plant conditions. The methodology The performance of SRV testing is not a Boiler and Pressure Vessel Code, with the for determining leakage provides for large precursor to any accident previously exception of opening the valve following margins between calculated and actual evaluated and does not change the manner in installation or maintenance for which a relief leakage values in the proposed limited which the valves are operated. The proposed request has been submitted (Ref. 6.1 [of the tubesheet inspection depth criteria. testing requirements will not contribute to September 29, 2011, application]), proposing Therefore, the proposed change does not the failure of the SRVs nor any plant an acceptable alternative. No setpoints are involve a significant reduction in any margin structure, system, or component. NextEra being changed which would alter the of safety. Energy Duane Arnold has determined that dynamic response of plant equipment.

The NRC staff has reviewed the the proposed change in testing methodology Accordingly, no new failure modes are srobinson on DSK4SPTVN1PROD with NOTICES licensees analysis and, based on this provides an equivalent level of assurance that introduced.

review, it appears that the three the SRVs are capable of performing their Therefore, the proposed change does not intended safety functions. Thus, the create the possibility of a new or different standards of 10 CFR 50.92(c) are proposed changes do not affect the kind of accident from any accident satisfied. Therefore, the NRC staff probability of an accident previously previously evaluated.

proposes to determine that the evaluated. 3. Does the proposed amendment involve amendment request involves no The performance of SRV testing provides a significant reduction in the margin of significant hazards consideration. confidence that the relief valves are capable safety?

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35076 Federal Register / Vol. 77, No. 113 / Tuesday, June 12, 2012 / Notices Response: No. consequences of an accident previously Therefore, the proposed change does not Overpressure protection of the RCPB evaluated? involve a significant reduction in a margin of

[reactor coolant pressure boundary] is based Response: No. safety.

on the SRVs setpoints and total relief As indicated in FSAR (plant-specific DCD) capacity. The setpoints are verified at an Subsection 3.8.5.5, the design function of the The NRC staff has reviewed the offsite testing facility; this requirement is not basemat is to provide the interface between licensees analysis and, based on this altered by the proposed change. The relief the nuclear island structures and the review, it appears that the three capacity of each SRV is determined by the supporting soil or rock. The basemat transfers standards of 10 CFR 50.92(c) are valves geometry, which is also not altered by the load of nuclear island structures to the satisfied. Therefore, the NRC staff the proposed test methods. supporting soil or rock. The basemat proposes to determine that the The proposed changes will allow testing of transmits seismic motions from the amendment request involves no the valve actuation electrical circuitry, supporting soil or rock to the nuclear island.

including the solenoid, and mechanical significant hazards consideration.

The revision of the basemat construction Attorney for licensee: Mr. M. Stanford actuation components, without causing the tolerance does not have an adverse impact on SRV to open. The SRVs will be manually Blanton, Balch & Bingham LLP, 1710 the response of the basemat and nuclear actuated prior to installation in the plant.

island structures to safe shutdown Sixth Avenue North, Birmingham, AL Therefore, all modes of SRV operation will be 35203-2015.

earthquake ground motions or loads due to tested prior to entering the mode of operation NRC Branch Chief: Mark E. Tonacci.

requiring the valves to perform their safety anticipated transients or postulated accident functions. The proposed changes do not conditions. The revision of the basemat Virginia Electric and Power Company, affect the valve setpoint or the operational construction tolerance does not impact the Docket No. 50-338 and 50-339, North criteria that cause the SRVs to open during support, design, or operation of mechanical and fluid systems. There is no change to Anna Power Station, Units 1 and 2, plant transients or accidents, either manually Louisa County, Virginia or automatically. There are no changes plant systems or the response of systems to proposed which alter the setpoints at which postulated accident conditions. There is no Date of amendment request: April 2, protective actions are initiated, and there is change to the predicted radioactive releases 2012.

no change to the operability requirements for due to normal operation or postulated Description of amendment request:

equipment assumed to operate for accident accident conditions. The plant response to The proposed amendment would delete mitigation. previously evaluated accidents or external Therefore, the proposed change does not events is not adversely affected, nor does the the Steam Generator Water Level Low involve a significant reduction in a margin of change described create any new accident Coincident with Steam Flow/Feedwater safety. precursors. Flow Mismatch Reactor Trip Function Therefore, there is no significant increase from the Technical Specification (TS)

The NRC staff has reviewed the in the probability or consequences of an Table 3.3.1-1 Item 15.

licensees analysis and, based on this accident previously evaluated. Basis for proposed no significant review, it appears that the three 2. Does the proposed amendment create hazards consideration determination:

standards of 10 CFR 50.92(c) are the possibility of a new or different kind of As required by 10 CFR 50.91(a), the satisfied. Therefore, the NRC staff accident from any accident previously licensee has provided its analysis of the proposes to determine that the evaluated?

issue of no significant hazards amendment request involves no Response: No.

The proposed change is to increase the consideration, which is presented significant hazards consideration.

construction tolerance for the basemat below:

Attorney for licensee: Mr. Mitchell S.

Ross, P. O. Box 14000 Juno Beach, FL thickness. The revision of the basemat Criterion 1Does the change involve a 33408-0420. construction tolerance does not change the significant increase in the probability or NRC Acting Branch Chief: Istvan design of the basemat or nuclear island consequences of an accident previously structures. The revision of the basemat evaluated?

Frankl.

construction tolerance does not change the The initiating conditions and assumptions Southern Nuclear Operating Company, design function, support, design, or operation for accidents described in the Updated Final Inc. Docket Nos.52-025 and 52-026, of mechanical and fluid systems. The Safety Analyses Report remain as previously Vogtle Electric Generating Plant (VEGP) revision of the basemat construction analyzed. The proposed change does not Units 3 and 4, Burke County, Georgia tolerance does not result in a new failure introduce a new accident initiator nor does mechanism for the basemat or new accident it introduce changes to any existing accident Date of amendment request: April 6, precursors. As a result, the design function initiators or scenarios described in the 2012, and revised on April 12 and May of the basemat is not adversely affected by Updated Final Safety Analyses Report. The 7, 2012. the proposed change. Steam/Feedwater Flow Mismatch and Low Description of amendment request: Therefore, the proposed change will not Steam Generator Water Level reactor trip is The proposed changes would amend create the possibility of a new or different not credited for accident mitigation in any Combined License Nos. NPF-91 and kind of accident from any accident accident analyses described in the Updated NPF-92 for Vogtle Electric Generating previously evaluated. Final Safety Analyses Report. The Steam/

3. Does the proposed amendment involve Feedwater Flow Mismatch and Low Steam Plant (VEGP) Units 3 and 4, a significant reduction in a margin of safety? Generator Water Level trip was designed to respectively, in regard to the upper Response: No. meet the control and protection systems tolerance on the Nuclear Island (NI) The revision in the basemat thickness interaction criteria of IEEE-279. The Steam critical sections basemat thickness as construction tolerance does not have an Generator Level Median Signal Selector identified in the plant-specific Design adverse impact on the strength of the (MSS) prevents adverse control and Control Document (DCD). basemat. The increase in the basemat protection system interaction such that it Basis for proposed no significant thickness construction tolerance does not replaces the need for the Steam/Feedwater hazards consideration determination: have an adverse impact on the seismic design Flow Mismatch and Low Steam Generator srobinson on DSK4SPTVN1PROD with NOTICES As required by 10 CFR 50.91(a), the spectra or the structural analysis of the Water Level reactor trip to satisfy the IEEE-licensee has provided its analysis of the basemat or other nuclear island structures. 279 requirements. As such, the affected issue of no significant hazards The revision in the basemat thickness control and protection systems will continue construction tolerance has no impact of the to perform their required functions without consideration, which is presented analysis of the nuclear island for sliding or adverse interaction, and maintain the below: overturning. As a result, the design function capability to shut down the reactor when

1. Does the proposed amendment involve of the basemat is not adversely affected by required on Low-Low Steam Generator water a significant increase in the probability or the proposed change. level. The ability to mitigate a loss of heat VerDate Mar<15>2010 22:42 Jun 11, 2012 Jkt 226001 PO 00000 Frm 00144 Fmt 4703 Sfmt 4703 E:\FR\FM\12JNN1.SGM 12JNN1

Federal Register / Vol. 77, No. 113 / Tuesday, June 12, 2012 / Notices 35077 sink accident previously evaluated is The MSS prevents adverse control and Criterion 1Does the change involve a unaffected. The frequency categories of protection system interaction such that it significant increase in the probability or previously evaluated accidents are not replaces the need for the Steam/Feedwater consequences of an accident previously changed. Flow Mismatch and Low Steam Generator evaluated?

Therefore, neither the probability of Water Level reactor trip and satisfies the The proposed change provides a new occurrence nor the consequences of an IEEE-279 requirements. Condition for two demand position accident previously evaluated is significantly indicators inoperable in one or more banks.

The proposed change improves the margin increased. The inoperability of two demand position of safety since removal of the Steam/

Criterion 2Does the change create the Feedwater Flow Mismatch and Low Steam indicators in one or more banks does not possibility of a new or different kind of Generator Water Level trip function directly affect any accident analysis or design accident from any accident previously basis limits or cause any limit not to be met, decreases the potential for challenges to plant evaluated? because the demand position indicator only safety systems, decreases the plant provides the intended demand as determined The substitution of the MSS for the Steam/ surveillance/maintenance activity, and Feedwater Flow Mismatch and Low Steam by the rod control system. The actual reduces plant complexity. These changes position of the control rods is determined by Generator Water Level trip will not introduce result in a reduction in the potential for use of the Rod Position Indications (RPIs) for any new failure modes to the required protection functions. The MSS only interacts unnecessary plant transients. each control rod, or the movable incore with the feedwater control system. The The Technical Specifications continue to detector system when the RPIs are Steam Generator Water Level Low-Low assure that the applicable operating inoperable.

protection function is not affected by this parameters and systems are maintained The inoperability of the demand position change. Isolation devices upstream of the within the design requirements and safety indicators does prevent the comparison of MSS circuitry ensure that the Steam analysis assumptions. Therefore, the the RPIs to the demand position indication Generator Water Level Low-Low protection elimination of this trip function will not for verification of rod insertion and rod group function is not affected. The MSS is designed alignment limits, which is conducted as a result in a significant reduction in the margin to reduce the frequency of system failures periodic surveillance to maintain the reactor of safety as defined in the Updated Final within analyzed conditions. The use of a 4 through utilization of highly reliable Safety Analyses Report or Technical components in a configuration that relies on hour Completion Time limit provides a Specifications. restriction that limits the time that reactor a minimum of additional equipment.

Therefore, it is concluded that this change operation can continue during this loss of the Components used in the MSS are of a quality consistent with low failure rates and does not involve a significant reduction in demand position indication. Since the loss of minimum maintenance requirements, and the margin of safety. the demand position indication does not conform to protection system requirements. cause the rods to change position, hence the Furthermore, the design provides the The NRC staff has reviewed the actual control rod positions are expected to capability for complete unit testing that licensees analysis and, based on this remain within required limits. Placing the provides unambiguous determination of review, it appears that the three Rod Control System in a condition incapable credible system failures. It is through these standards of 50.92(c) are satisfied. of rod movement is a positive control to features that the overall design of the MSS Therefore, the NRC staff proposes to prevent rod stepping while maintenance is minimizes the occurrence of undetected determine that the amendment request being performed.

failures that may exist between test intervals. The proposed change to allow two demand involves no significant hazards Therefore, the possibility for a new or position indicators to be inoperable in one or consideration. more banks does not affect the automatic or different kind of accident from any accident previously evaluated is not created. Attorney for licensee: Lillian M. manual shutdown capability of the reactor Cuoco, Senior Counsel, Dominion protection system and no accident analyses Criterion 3Does this change involve a Resources Services, Inc., 120 Tredegar are impacted by the proposed change. The significant reduction in a margin of safety?

Street, RS-2, Richmond, VA 23219. operability of the control rods is not affected The proposed amendment does not involve by the inoperability of the demand position revisions to any safety analysis limits or NRC Branch Chief: Nancy L. Salgado. indicators.

safety system settings that will adversely Therefore, neither the probability of impact plant safety. The proposed Virginia Electric and Power Company, occurrence nor the consequences of an amendment does not alter the functional Docket No. 50-339, North Anna Power accident previously evaluated is significantly capabilities assumed in a safety analysis for Station, Unit 2, Louisa County, Virginia increased.

any system, structure, or component important to the mitigation and control of Date of amendment request: May 11, Criterion 2Does the change create the design bases accident conditions within the possibility of a new or different kind of 2012. accident from any accident previously facility. Nor does this amendment revise any parameters or operating restrictions that are Description of amendment request: evaluated?

assumptions of a design basis accident. In The proposed amendment would revise The proposed change provides new addition, the proposed amendment does not the Technical Specification (TS) 3.1.7, requirements for two demand position affect the ability of safety systems to ensure Rod Position Indication to allow two indicators inoperable in one or more banks.

that the facility can be placed and demand position indicators in one or No new accident initiators are introduced by maintained in a shutdown condition for more banks to be inoperable for up to the proposed requirements because the extended periods of time. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This change is proposed as a allowed condition for inoperability of the The ability of the Steam Generator Water temporary change to the TS for the demand position indicators does not cause Level Low-Low reactor trip function credited any new failure modes to be created that can in the safety analysis to protect against a current operating cycle and is proposed cause an accident. The proposed change does sudden loss of heat sink event is not affected as a footnote to the current TS Limiting not affect the reactor protection system or the by the proposed change: Since the Steam Condition for Operation (LCO) Section reactor control system. The control rods Generator Low-Low Level trip is credited 3.1.7, Condition D. should remain within the required limits srobinson on DSK4SPTVN1PROD with NOTICES alone as providing complete protection for Basis for proposed no significant because the failure of the demand position the accident transients that result in low indicators does not cause the rods to change hazards consideration determination:

steam generator level, eliminating the Steam/ position and the RPIs remain available in the Feedwater Flow Mismatch and Low Steam As required by 10 CFR 50.91(a), the affected banks to verify the position of the Generator Water Level trip will not change licensee has provided its analysis of the control rods. In addition, the Rod Control any safety analysis conclusion for any issue of no significant hazards System is placed in a condition incapable of analyzed accident described in the Updated consideration, which is presented rod movement as a positive control to Final Safety Analyses Report. below: prevent rod stepping while maintenance is VerDate Mar<15>2010 22:42 Jun 11, 2012 Jkt 226001 PO 00000 Frm 00145 Fmt 4703 Sfmt 4703 E:\FR\FM\12JNN1.SGM 12JNN1

35078 Federal Register / Vol. 77, No. 113 / Tuesday, June 12, 2012 / Notices being performed. Hence, no new failure SourcesOperating, Surveillance requires the diesel generators to be tested at modes or accident sequences are created that Requirements related to Diesel increased loads which envelope the actual would cause a new or different kind of Generator test loads, voltage, and power demand requirements for the diesel accident from any accident previously generators during design basis conditions.

frequency. The proposed changes will These revised loads continue to demonstrate evaluated.

Therefore, the possibility for a new or correct non-conservative Diesel the capability and capacity of the diesel different kind of accident from any accident Generator load values that are currently generators to perform their required previously evaluated is not created. under administrative controls. functions. There are no new failure modes or Criterion 3Does this change involve a Basis for proposed no significant mechanisms created due to testing the diesel significant reduction in a margin of safety? hazards consideration determination: generators at the proposed test loading.

As required by 10 CFR 50.91(a), the Testing of the emergency diesel generators at The operability of the RPIs is required to the proposed test loadings does not involve determine control rod positions and thereby licensee has provided its analysis of the any modification in the operational limits or ensure compliance with the control rod issue of no significant hazards physical design of plant systems. There are alignment and insertion limits. The proposed consideration, which is presented no new accident precursors generated due to change does not alter the requirement to below: the proposed test loadings.

determine rod position, but provides a new 1. Does the proposed change involve a Therefore, it is concluded that the Condition for two demand position significant increase in the probability or proposed change does not create the indicators inoperable in one or more banks. consequences of an accident previously possibility of a new or different kind of The inoperability of two demand position evaluated? accident from any accident previously indicators for one or more banks results in Response: No. evaluated.

the reduced ability to periodically verify that The diesel generators are required to be 3. Does the proposed change involve a RPIs are operable and within expected limits. OPERABLE in the event of a design basis significant reduction in a margin of safety?

The condition does prevent the comparison accident coincident with a loss of offsite Response: No.

of the RPIs to the demand position indication power to mitigate the consequences of the The proposed Technical Specification for verification of rod insertion and rod group accident. The diesel generators are not change will continue to demonstrate that the alignment limits, which is conducted as accident initiators and therefore these diesel generators meet the Technical periodic surveillance to maintain the reactor changes do not involve a significant increase Specification definition of OPERABILITY, within analyzed conditions. The loss of the in the probability of an accident previously that is, the proposed tests will demonstrate demand position indication does not cause evaluated. that the diesel generators will perform their the rods to change position, hence the actual The accident analyses assume that at least safety function and the necessary diesel control rod positions are expected to remain one engineered safety feature bus is provided generator attendant instrumentation, within required limits. The use of a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> controls, cooling, lubrication and other with power either from the offsite circuits or Completion Time limit provides a restriction auxiliary equipment required for the the diesel generators. The Technical that limits the time that reactor operation can emergency diesel generators to perform their Specification change proposed in this license continue during this loss of the demand safety function loads are also tested at these amendment request will continue to assure position indication. This ensures the proposed loadings. The proposed testing will that the diesel generators have the capacity condition is promptly corrected or the reactor also continue to demonstrate the capability and capability to assume their maximum shutdown in accordance with the applicable and capacity of the diesel generators to design basis accident loads. The proposed Technical Specifications action statements. supply their required loads for mitigating a change does not significantly change how the Thus, the proposed change maintains the design basis accident.

operation of the reactor within the applicable plant would mitigate an accident previously evaluated. The proposed change does not alter the margins of safety because the inoperability manner in which safety limits, limiting safety will be corrected or the unit will be The proposed change does not adversely affect accident initiators or precursors nor system settings or limiting conditions for shutdown prior to any significant reduction operation are determined. The safety analysis in the ability to verify control rod position by alter the design assumptions, conditions, and configuration of the facility or the manner in acceptance criteria are not impacted by this the use analog RPIs. change. The proposed change will not result Therefore, it is concluded that this change which the plant is operated and maintained.

The proposed change does not adversely in plant operation in a configuration outside does not involve a significant reduction in the design basis.

the margin of safety. affect the ability of structures, systems, and components (SSC) to perform their intended Therefore, it is concluded that the The NRC staff has reviewed the safety function to mitigate the consequences proposed change does not involve a significant reduction in a margin of safety.

licensees analysis and, based on this of an initiating event within the assumed review, it appears that the three acceptance limits. The proposed change does The NRC staff has reviewed the standards of 50.92(c) are satisfied. not affect the source term, containment licensees analysis and, based on this Therefore, the NRC staff proposes to isolation, or radiological release assumptions review, it appears that the three used in evaluating the radiological standards of 10 CFR 50.92(c) are determine that the amendment request consequences of any accident previously involves no significant hazards evaluated. Further, the proposed change does satisfied. Therefore, the NRC staff consideration. not increase the types and amounts of proposes to determine that the Attorney for licensee: Lillian M. radioactive effluent that may be released amendment request involves no Cuoco, Senior Counsel, Dominion offsite, nor significantly increase individual significant hazards consideration.

Resources Services, Inc., 120 Tredegar or cumulative occupational/public radiation Attorney for licensee: Jay Silberg, Esq.,

Street, RS-2, Richmond, VA 23219. exposure. Pillsbury Winthrop Shaw Pittman LLP, NRC Branch Chief: Nancy L. Salgado. Therefore, the proposed change does not 2300 N Street NW., Washington, DC represent a significant increase the 20037.

Wolf Creek Nuclear Operating probability or consequences of an accident NRC Branch Chief: Michael T.

Corporation, Docket No. 50-482, Wolf previously evaluated. Markley.

Creek Generating Station, Coffey 2. Does the proposed change create the srobinson on DSK4SPTVN1PROD with NOTICES County, Kansas possibility of a new or different kind of Notice of Issuance of Amendments to accident from any accident previously Facility Operating Licenses and Date of amendment request: evaluated?

November 30, 2011. Combined Licenses Response: No.

Description of amendment request: The proposed Technical Specification During the period since publication of The proposed amendment would revise change does not involve a change in the plant the last biweekly notice, the the Wolf Creek Generating Station design, system operation, or the use of the Commission has issued the following Technical Specification (TS) 3.8.1, AC diesel generators. The proposed change amendments. The Commission has VerDate Mar<15>2010 22:42 Jun 11, 2012 Jkt 226001 PO 00000 Frm 00146 Fmt 4703 Sfmt 4703 E:\FR\FM\12JNN1.SGM 12JNN1

Federal Register / Vol. 77, No. 113 / Tuesday, June 12, 2012 / Notices 35079 determined for each of these revise the allowable value (AV) and Date of initial notice in Federal amendments that the application related setpoints for the Main Steam Register: September 6, 2011 (76 FR complies with the standards and Tunnel Temperature functions 1.e, 3.f, 55129).

requirements of the Atomic Energy Act and 4.h in TS Table 3.3.6.1-1. In The letter dated November 10, 2011, of 1954, as amended (the Act), and the addition, the RBSs Emergency Action provided clarifying information that did Commissions rules and regulations. Levels will be revised to reflect the not change the initial proposed no The Commission has made appropriate changes to the AV and related setpoints significant hazards consideration findings as required by the Act and the in TS 3.3.6.1. determination or expand the application Commissions rules and regulations in Date of issuance: May 30, 2012. beyond the scope of the original Federal 10 CFR Chapter I, which are set forth in Effective date: As of the date of Register notice.

the license amendment. issuance and shall be implemented The Commissions related evaluation A notice of consideration of issuance within 60 days from the date of of the amendments is contained in a of amendment to facility operating issuance. Safety Evaluation dated May 25, 2012.

license or combined license, as Amendment No.: 174. No significant hazards consideration applicable, proposed no significant Facility Operating License No. NPF- comments received: No.

hazards consideration determination, 47: The amendment revised the Facility and opportunity for a hearing in South Carolina Electric and Gas Operating License and Technical Company, Docket No. 50-395, Virgil C.

connection with these actions, was Specifications.

published in the Federal Register as Summer, Nuclear Station (VCSNS), Unit Date of initial notice in Federal 1, Jenkinsville, South Carolina indicated. Register: February 7, 2012 (77 FR 6147).

Unless otherwise indicated, the Date of application for amendment:

The supplemental letters dated Commission has determined that these August 11, 2011.

September 16, 2011, and February 7, amendments satisfy the criteria for Brief description of amendment: This February 24, and April 3, 2012, categorical exclusion in accordance amendment revised the VCSNS provided additional information that with 10 CFR 51.22. Therefore, pursuant Technical Specification (TS) to allow an clarified the application, did not expand to 10 CFR 51.22(b), no environmental updating of the applicable topical report the scope of the application as originally impact statement or environmental in TS 6.9.1.11, Core Operating Limits noticed, and did not change the staffs assessment need be prepared for these Report to use the three-dimensional amendments. If the Commission has original proposed no significant hazards consideration determination as Advanced Nodal Code neutronic model.

prepared an environmental assessment Date of Issuance: May 30, 2012.

under the special circumstances published in the Federal Register.

The Commissions related evaluation Effective date: As of the date of provision in 10 CFR 51.22(b) and has issuance and shall be implemented made a determination based on that of the amendment is contained in a Safety Evaluation dated May 30, 2012. within 90 days.

assessment, it is so indicated. Amendment No: 190.

For further details with respect to the No significant hazards consideration comments received: No. Renewed Facility Operating License action see (1) the applications for No. NPF-12: Amendment revises the amendment, (2) the amendment, and (3) Exelon Generation Company, LLC, and License and Technical Specifications.

the Commissions related letter, Safety PSEG Nuclear LLC, Docket Nos. 50-277 Date of initial notice in Federal Evaluation and/or Environmental and 50-278, Peach Bottom Atomic Register: October 11, 2011 (76 FR Assessment as indicated. All of these Power Station, Units 2 and 3, York and 62864).

items are available for public inspection Lancaster Counties, Pennsylvania The Commissions related evaluation at the NRCs Public Document Room of the amendment is contained in a (PDR), located at One White Flint North, Date of application for amendments:

June 2, 2011, as supplemented by letter Safety Evaluation dated May 30, 2012.

Room O1-F21, 11555 Rockville Pike No significant hazards consideration (first floor), Rockville, Maryland 20852. dated November 10, 2011.

Brief description of amendments: The comments received: No.

Publicly available documents created or received at the NRC are available online amendments modify Technical Dated at Rockville, Maryland, this 1st day in the NRC Library at http:// Specification (TS) 3.1.2, Reactivity of June, 2012.

Anomalies, to change the method used For the Nuclear Regulatory Commission.

www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if to perform the reactivity anomaly Michele G. Evans, there are problems in accessing the surveillance. Specifically, the Director, Division of Operating Reactor documents located in ADAMS, contact amendments allow performance of the Licensing, Office of Nuclear Reactor surveillance based on the difference Regulation.

the PDRs Reference staff at 1-800-397-4209, 301-415-4737 or by email to between the monitored (i.e., actual) core [FR Doc. 2012-13921 Filed 6-11-12; 8:45 am]

pdr.resource@nrc.gov. reactivity and the predicted core BILLING CODE 7590-01-P reactivity. The surveillance was Entergy Gulf States Louisiana, LLC, and previously performed based on the Entergy Operations, Inc., Docket No. 50- difference between the monitored NUCLEAR REGULATORY 458, River Bend Station, Unit 1, West control rod density and the predicted COMMISSION Feliciana Parish, Louisiana control rod density.

[Docket No. 50-443; NRC-2010-0206]

Date of amendment request: July 27, Date of issuance: May 25, 2012.

2011, as supplemented by letters dated Effective date: As of the date of srobinson on DSK4SPTVN1PROD with NOTICES License Renewal Application for September 16, 2011, and February 7, issuance, to be implemented within 60 Seabrook Station, Unit 1 ; NextEra February 24, and April 3, 2012. days. Energy Seabrook, LLC Brief description of amendment: The Amendments Nos.: 284 and 287.

amendment modified River Bend Renewed Facility Operating License AGENCY: Nuclear Regulatory Stations (RBS) Technical Specification Nos. DPR-44 and DPR-56: The Commission.

(TS) 3.3.6.1, Primary Containment and amendments revised the License and ACTION: License renewal application; Drywell Isolation Instrumentation, to TSs. intent to prepare supplement to draft VerDate Mar<15>2010 22:42 Jun 11, 2012 Jkt 226001 PO 00000 Frm 00147 Fmt 4703 Sfmt 4703 E:\FR\FM\12JNN1.SGM 12JNN1