Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations
Federal Register: June 29, 2010 (Volume 75, Number 124)
Notices
Page 37471-37478
From the Federal Register Online via GPO Access [wais.access.gpo.gov]
DOCID:fr29jn10-101
NUCLEAR REGULATORY COMMISSION
NRC-2010-0232
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
-
Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or proposed to be issued from June 3, 2010 to June 16, 2010. The last biweekly notice was published on June 15, 2010 (75 FR 33839).
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 Cindy Bladey,
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
O1F21, 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.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/ reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or a presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address, and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also identify the specific contentions which the requestor/
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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, 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 http://www.nrc.gov/site-help/e- submittals/apply-certificates.html. System requirements for accessing the E-Submittal server are detailed in NRC's ``Guidance for Electronic
Submission,'' which is available on the agency's public Web site at http://www.nrc.gov/site-help/e-submittals.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Meta System Help Desk will not be able to offer assistance in using unlisted software.
If a participant is electronically submitting a document to the NRC in accordance with the E-Filing rule, the participant must file the document using the NRC's online, Web-based submission form. In order to serve documents through EIE, users will be required to install a Web browser plug-in from the NRC Web site. Further information on the Web- based submission form, including the installation of the Web browser plug-in, is available on the NRC's public Web site at http:// www.nrc.gov/site-help/e-submittals.html.
Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a request for hearing or petition for leave to intervene. Submissions should be in
Portable Document Format (PDF) in accordance with NRC guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e- submittals.html. A filing is considered complete at the time the documents are submitted through the NRC's E-Filing system. To be timely, an electronic filing must be submitted to the E-Filing system no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an e-mail notice confirming receipt of the document. The
E-Filing system also distributes an e-mail notice that provides access to the document to the NRC Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing request/petition to intervene is filed so that they can obtain access to the document via the E-Filing system.
A person filing electronically using the agency's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System
Help Desk through the ``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html, by e-mail at
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
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document on all other participants. Filing is considered complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists.
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at http://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission.
Petitions for leave to intervene must be filed no later than 60 days from the date of publication of this notice. Non-timely filings will not be entertained absent a determination by the presiding officer that the petition or request should be granted or the contentions should be admitted, based on a balancing of the factors specified in 10
CFR 2.309(c)(1)(i)-(viii).
For further details with respect to this license amendment application, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint
North, Public File Area O1F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from the ADAMS Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to ADAMS or who encounter problems in accessing the documents located in ADAMS, should contact the NRC PDR Reference staff at 1-800- 397-4209, 301-415-4737, or by e-mail to pdr.resource@nrc.gov.
Dominion Energy Kewaunee, Inc. Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of amendment request: April 13, 2010.
Description of amendment request: The licensee proposed to revise
Section 3.1.a.1.C, ``Reactor Coolant Pumps,'' Section 3.1.a.3,
``Pressurizer Safety Valves,'' and Section 3.1.b, ``Heatup and Normal
Cooldown Limit Curves for Normal Operation,'' of the Technical
Specifications (TS). Specifically, the proposed amendment would replace the heatup and cooldown pressure-temperature (P-T) limit curves with new ones, and specifying a higher low temperature overpressure protection (LTOP) enabling temperature.
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 (NSHC) analysis. The NRC staff reviewed the licensee's
NSHC analysis and has prepared its own as follows: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The purpose of the P-T limits curves and LTOP is to ensure that the reactor vessel is operated within its material design limits. As such, the subject specifications specify the pressure limits inside the reactor vessel under different temperature conditions for normal operation. No conditions of operation within the approved P-T limits were postulated to be initiators of accidents previously analyzed in the Kewaunee Final Safety Analysis Report. Furthermore, the consequences of the analyzed accidents were not postulated to be exacerbated by normal operation within approved P-T limits.
Accordingly, the probability of occurrence and the consequences of the previously analyzed accidents would not be affected in any way by the proposed P-T limits changes. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve any physical alteration of the plant (no new or different type of equipment will be installed) nor does it change methods and procedures governing plant operation.
The proposed change will not impose any new or eliminate any old requirements. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed change will not reduce a margin of safety because it has no effect on any safety analysis methods, scenarios, or assumptions. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
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 proposed amendment involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Resources Services, Inc., Counsel for Dominion Energy Kewaunee, Inc., 120 Tredegar Street, Richmond, VA 23219.
NRC Branch Chief: Robert J. Pascarelli.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: April 28, 2010.
Description of amendment request: The proposed change revises the
Final Safety Analysis Report and Emergency Plan to support U.S.
Department of Energy non-intrusive surveillance and characterization activities within the 618-11 Waste Burial Ground.
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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
Normal and postulated activities at the 618-11 site do not serve as initiators of any Columbia [Generating Station] accident previously evaluated, nor do they require reassessment of the previously evaluated accidents. The accident probabilities are unaffected and the outcomes remain unchanged.
Therefore there is no 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 analyzed?
Response: No.
The only hazard postulated beyond the 618-11 site and onto the
Columbia facility is a release of 44.5 mrem [millirem] at 100 m
meters
. This level of exposure does not impact the design function or operation of any Columbia SSCs [structures, systems, or components]. The protected area of the facility that encloses the safety related SSCs is greater than 300 m from the postulated release point. The calculated dose at 300 m is 3 mrem. This level of exposure does not cause any new or different kind of accident.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
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(3) Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The only hazard postulated beyond the 618-11 site and onto the
Columbia facility is a release of 44.5 mrem at 100 m. This level of exposure does not impact the design function or operation of any
Columbia SSCs. The protected area of the facility that encloses the safety related SSCs is greater than 300 m from the postulated release point. The calculated dose at 300 m is 3 mrem. This level of exposure does not impact the equipment qualification of SSCs and is well within the mild environment range for SSCs. It does not exceed or alter a design safety limit.
Therefore, the proposed change does not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: William A. Horin, Esq., Winston & Strawn, 1700 K Street, NW., Washington, DC 20006-3817.
NRC Branch Chief: Michael T. Markley.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: April 13, 2010.
Description of amendment request: The proposed amendment would revise Technical Specification (TS) to institute a requirement to perform a Logic System Functional Test of the Control Rod Block actuation instrumentation trip functions once every 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: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The change does not impact the function of any structure, system or component that affects the probability of an accident or that supports mitigation or consequences of an accident previously evaluated. The proposed change adds a requirement to perform additional testing of the control rod block instrumentation. The proposed change does not affect reactor operations or accident analysis and there is no change to the radiological consequences of a previously analyzed accident. The operability requirements for accident mitigation systems remain consistent with the licensing and design basis.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve any physical alteration of plant equipment and does not change the method by which any safety- related system performs its function. The proposed change involves the addition of a requirement to perform a logic system functional test of plant instrumentation. This test is within the design capability of the system and does not create the possibility of a different kind of accident. No new or different types of equipment will be permanently installed. Operation of existing installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
These changes do not change any existing design or operational requirements and do not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. The proposed change only affects the testing of the control rod block instrumentation. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed change.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Nancy Salgado.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: April 13, 2010.
Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to update the Table of
Contents and the Applicability and Objective portions of TS 4.12 as a result of changes made by License Amendments 230 and 239, and to revise wording in TS 3.7.A.8. The proposed changes are considered administrative in nature and do not materially change any technical requirement.
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. The operation of Vermont Yankee Nuclear Power Station (VY) in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed changes are administrative in nature and do not involve any physical changes to the plant. The changes do not revise the methods of plant operation which could increase the probability or consequences of accidents. No new modes of operation are introduced by the proposed changes such that a previously evaluated accident is more likely to occur or more adverse consequences would result.
Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. The operation of VY in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes are administrative in nature and do not affect the operation of any systems or equipment, nor do they involve any potential initiating events that would create any new or different kind of accident. There are no changes to the design assumptions, conditions, configuration of the facility, or manner in which the plant is operated and maintained. The changes do not affect assumptions contained in plant safety analyses or the physical design and/or modes of plant operation. Consequently, no new failure mode is introduced.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated. 3. The operation of VY in accordance with the proposed amendment will not involve a significant reduction in a margin of safety.
There are no changes being made to the Technical Specification
(TS) safety limits or safety system settings. The operating limits and functional capabilities of systems, structures and components are unchanged as a result of these administrative changes. These changes do not affect any equipment involved in potential initiating events or plant response to accidents. There is no change to the basis for any TS related to the establishment, or maintenance of, a nuclear safety margin. The proposed changes do not impact any safety limits, safety settings or safety margins.
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Therefore, operation of VY in accordance with the proposed amendment will not involve a significant reduction in the margin to 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: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Nancy Salgado.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: March 31, 2010.
Description of amendment request: The proposed amendment would implement an alternative source term (AST) for Arkansas Nuclear One,
Unit 2 (ANO-2). The proposed amendment would modify Technical
Specification (TS) 3.4.8, ``Specific Activity,'' and 6.5.12, ``Control
Room Habitability Program,'' and associated definitions as related to the use of an AST associated with accident offsite and control room dose consequences.
Basis for proposed no significant hazards consideration determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR) Section 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 use of an AST is recognized in 10 CFR 50.67. RG [Regulatory
Guide] 1.183 provides guidance for implementation of an AST. The AST involves quantities, isotopic composition, chemical and physical characteristics, and release timing of radioactive material for use as inputs to accident dose analyses. As such, the AST cannot affect the probability of occurrence of a previously evaluated accident. In addition, the increase in the DEX [Dose Equivalent Xenon-133] activity limit and the terminology/reference changes proposed for the ANO-2 TSs are unrelated to accident initiators. No facility equipment, procedure, or process changes are required in conjunction with implementing the AST that could increase the likelihood of a previously analyzed accident. The proposed changes in the source term and the methodology for the dose consequence analyses follow the guidance of RG 1.183. As a result, there is no increase in the likelihood of existing event initiators.
Regarding accident consequences, the increase in the DEX activity limit acts to support the analysis results given the application of an AST. The proposed limit was utilized as an assumption in the AST analysis and determined to be acceptable. The results of accident dose analyses using the AST are compared to TEDE
Total Effective Dose Equivalent
acceptance criteria that account for the sum of deep dose equivalent (for external exposure) and committed effective dose equivalent (for internal exposure). Dose results were previously compared to separate limits on whole body, thyroid, and skin doses as appropriate for the particular accident analyzed. The results of the revised dose consequences analyses demonstrate that the regulatory acceptance criteria are met for each analyzed event. The proposed TS terminology/reference changes are consistent with the analysis and adoption of an AST. Implementing the AST involves no facility equipment, procedure, or process changes that could affect the radioactive material actually released during an event. Subsequently, no conditions have been created that could significantly increase the consequences of any of the events being evaluated.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of any of the events being 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 AST involves quantities, isotopic composition, chemical and physical characteristics, and release timing of radioactive material for use as inputs to accident dose analyses. As such, the AST cannot create the possibility of a new or different kind of accident. In addition, the increased DEX activity limit and proposed terminology/ reference changes within the TSs are unrelated to accident initiators and are supported by AST adoption.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
Implementing the AST is relevant only to calculated accident dose consequences. The results of the revised dose consequences analyses demonstrate that the regulatory acceptance criteria are met for each analyzed event. In addition, the increased DEX activity limit and proposed terminology/reference changes within the TSs support adoption of the AST methodologies, have been determined to result in acceptable dose consequence and do not result in a significant impact to any margin of safety. The AST does not affect the transient behavior of non-radiological parameters (e.g., RCS
Reactor Coolant System
pressure, Containment pressure) that are pertinent to a margin of safety.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: Joseph A. Aluise, Associate General
Counsel--Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New
Orleans, Louisiana 70113.
NRC Branch Chief: Michael T. Markley.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: April 19, 2010.
Description of amendment request: The proposed amendments would revise Technical Specification 3.4.11, ``RCS Pressure and Temperature
(P/T) Limits,'' to incorporate revised P/T curves that are valid for up to 32 effective full power years of operation.
Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises Technical Specification (TS) Section 3.4.11 to replace the existing P/T curves with revised curves that are valid up to 32 EFPY. The revised curves were developed using the methodology of General Electric (GE) Topical Report NEDC-32983P,
``General Electric Methodology for Reactor Pressure Vessel Fast
Neutron Flux Evaluations.'' The NEDC-32983P methodology has been approved by the NRC for use by licensees. The P/T limits are not derived from design basis accident analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause non-ductile failure of the reactor coolant pressure boundary, a condition that is unanalyzed.
Since the P/T limits are not derived from any design basis accident, there are no acceptance limits related to the P/T limits. Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.
Thus, the proposed changes do not have any affect on the probability of an accident previously evaluated.
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The P/T curves are used as operational limits during heatup or cooldown maneuvering, when the pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region. The P/T curves provide assurance that station operation is consistent with a previously evaluated accident. Thus, the radiological consequences of any accident previously evaluated are not increased.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not change the response of plant equipment to transient conditions. The proposed change does not introduce any new equipment, modes of system operation, or failure mechanisms.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed change adopts P/T curves that have been developed using the methodology of GE Topical Report NEDC-32983P. The NEDC- 32983P methodology adheres to the guidance in NRC Regulatory Guide 1.190, ``Calculation and Dosimetry methods for Determining Pressure
Vessel Neutron Fluence,'' dated March 2001. In a letter dated
September 14, 2001, the NRC approved NEDC-32983P for use by licensees. The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The setpoints at which protective actions are initiated are not altered by the proposed change.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Stephen J. Campbell.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: March 29, 2010, as supplemented on May 28, 2010.
Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to extend the allowed outage time (AOT) for the ``A'' and ``B'' emergency diesel generators (EDGs) from 72 hours to 14 days.
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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The emergency diesel generators are safety related components which provide backup electrical power supply to the onsite
Safeguards Distribution System. The emergency diesel generators are not accident initiators; the EDGs are designed to mitigate the consequences of previously evaluated accidents including a loss of offsite power [LOOP]. Extending the AOT for a single EDG would not affect the previously evaluated accidents since the remaining EDGs supporting the redundant Engineered Safety Features (ESF) systems would continue to be available to perform the accident mitigation functions.
Thus allowing an emergency diesel generator to be inoperable for an additional 11 days for performance of maintenance or testing does not increase the probability of a previously evaluated accident.
Deterministic and probabilistic risk assessments evaluated the effect of the proposed Technical Specification changes on the availability of an electrical power supply to the plant emergency safeguards features systems. These assessments concluded that the proposed Technical Specification changes do not involve a significant increase in the risk of power supply unavailability.
There is incremental risk associated with continued operation for an additional 11 days with one emergency diesel generator inoperable; however, the calculated impact on risk is very small and is consistent with the acceptance guidelines contained in Regulatory
Guides 1.174 and 1.177. This risk is judged to be reasonably consistent with the risk associated with operations for 72 hours with one emergency diesel generator inoperable as allowed by the current Technical Specifications. Specifically, the remaining operable emergency diesel generators and paths are adequate to supply electrical power to the onsite Safeguards Distribution
System. An emergency diesel generator is required to operate only if both offsite power sources fail and there is an event which requires operation of the plant emergency safeguards features such as a design basis accident. The probability of a design basis accident occurring during this period is low.
The consequences of previously evaluated accidents will remain the same during the proposed 14 day AOT as during the current 72 hour AOT. The ability of the remaining TS required EDG to mitigate the consequences of an accident will not be affected since no additional failures are postulated while equipment is inoperable within the TS AOT. The standby power supply for each of the four safety-related load groups consists of one EDG complete with its auxiliaries, which include the cooling water, starting air, lubrication, intake and exhaust, and fuel oil systems. The sizing of the EDGs and the loads assigned among them is such that any combination of three out of four of these EDGs is capable of shutting down the plant safely, maintaining the plant in a safe shutdown condition, and mitigating the consequences of accident conditions.
Thus, this change does not involve a significant increase in the probability or consequences of a previously analyzed accident. 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed Technical Specification changes do not involve a change in the plant design, system operation, or procedures involved with the emergency diesel generators. The proposed changes allow an emergency diesel generator to be inoperable for additional time.
Equipment will be operated in the same configuration and manner that is currently allowed and designed for. There are no new failure modes or mechanisms created due to plant operation for an extended period to perform emergency diesel generator maintenance or testing.
Extended operation with an inoperable emergency diesel generator does not involve any modification in the operational limits or physical design of plant systems. There are no new accident precursors generated due to the extended AOT.
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.
Currently, if an inoperable emergency diesel generator is not restored to operable status within 72 hours, Technical Specification 3.8.1.1 ACTION b requires the unit be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. The proposed Technical Specification changes will allow steady state plant operation at 100% power for an additional 11 days.
Deterministic and probabilistic risk assessments evaluated the effect of the proposed Technical Specification changes on the availability of an electrical power supply to the plant emergency safeguards features systems. These assessments concluded that the proposed Technical Specification changes do not involve a significant increase in the risk of power supply unavailability.
The EDGs continue to meet their design requirements; there is no reduction in capability or change in design configuration. The EDG response to LOOP, LOCA [loss-of-
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coolant accident], SBO [station blackout], or fire is not changed by this proposed amendment; there is no change to the EDG operating parameters. In the extended AOT, as in the existing AOT, the remaining operable emergency diesel generators and paths are adequate to supply electrical power to the onsite Safeguards
Distribution System. The proposed change does not alter a design basis or safety limit; therefore it does not significantly reduce the margin of safety. The EDGs will continue to operate per the existing design and regulatory requirements.
Therefore, based on the considerations given above, 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: Vincent Zabielski, PSEG Nuclear LLC--N21,
P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: April 13, 2010, as supplemented by letter dated June 1, 2010.
Description of amendment request: The proposed amendment to Renewed
Facility Operating License No. NPF-42 would revise the approved fire protection program, as described in the Wolf Creek Generating Station
Updated Safety Analysis Report, by removing the high/low pressure interface designation of the pressurizer power-operated relief valves
(PORVs) and their associated block valves.
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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The design function of structures, systems and components are not impacted by the proposed change. This amendment classifies the pressurizer PORVs and their associated block valves based on the guidance in Regulatory Guide 1.189, ``Fire Protection for Nuclear
Power Plants,'' Revision 2, and Nuclear Energy [Institute] (NEI) 00- 01, ``Guidance for Post-Fire Safe-Shutdown Circuit Analysis,''
Revision 2, Appendix C. The classification change only affects the post fire safe shutdown (PFSSD) analysis methodology for the PORVs and block valves. Reclassification of the PORVs and block valves will not impact the use of the valves to depressurize the Reactor
Coolant System (RCS) to recover from certain transients if normal pressurizer spray is not available.
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.
There are no changes in the method by which any safety related plant system performs its safety function, and the normal manner of plant operation is unaffected. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of this change. There will be no adverse effect or challenges imposed on any safety related system as a result of this change. The classification change only affects the
PFSSD analysis methodology for the PORVs and block valves.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
There will be no effect on the manner in which safety limits or limiting safety system settings are determined nor will there be any effect on those plant systems necessary to ensure the accomplishment of protection functions. There will be no impact on departure from nuclear boiling [ratio] (DNBR) limits, heat flux hot channel factor
(FQ(Z)) limits, nuclear enthalpy rise hot channel factor
(F\N\[Delta]H) limits, peak centerline temperature (PCT) limits, peak local power density or any other margin of safety.
Therefore, this change does not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: Michael T. Markley.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing in connection with these actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.22(b) and has made a determination based on that assessment, it is so indicated.
For further details with respect to the action see (1) The applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection at the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http:// www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to pdr.resource@nrc.gov.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: August 17, 2009, as supplemented by letter dated January 21, 2010.
Brief description of amendment: The amendment modified (1)
Technical Specification (TS) 3.8.3, ``Diesel Fuel Oil, Lube Oil, and
Starting Air,'' to relocate specific numerical values for fuel oil and lube oil storage volumes
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from the TS to the TS Bases, (2) TS 3.8.1, ``AC [Alternating Current]
Sources--Operating,'' to relocate specific values for the day tank fuel oil volumes from the TS to the TS Bases, and (3) TS 5.5.9, ``Diesel
Fuel Oil Testing Program,'' to relocate the specific standard for particulate concentration testing of fuel oil from the TS to the TS
Bases.
Date of issuance: May 27, 2010.
Effective date: As of its date of issuance and shall be implemented within 90 days from the date of issuance.
Amendment No.: 215.
Facility Operating License No. NPF-21: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: November 3, 2009 (74 FR 56884). The supplemental letter dated January 21, 2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated May 27, 2010.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant, Van Buren County, Michigan
Date of application for amendment: March 31, 2010, supplemented by letter dated May 13, 2010.
Brief description of amendment: The amendment adds a new license condition 2.C (4) to Palisades Nuclear Plant, renewed facility license
No. DPR-20. This license condition would state that performance of
Technical Specification (TS) surveillance requirement (SR) 3.1.4.3 is not required for control rod drive 22 through cycle 21 or until the next entry into Mode 3. The amendment consists of changes to TS by addition of a note in SR 3.1.4.3, stating:
``Not required to be performed or met for control rod 22 during cycle 21 provided control rod 22 is administratively declared immovable, but trippable and Condition D is entered for control rod 22.''
Date of issuance: June 2, 2010.
Effective date: As of the date of issuance and shall be implemented within 15 days.
Amendment No.: 239.
Facility Operating License No. DPR-20: Amendment revised the
Technical Specifications and license.
Public comments requested as to proposed no significant hazards consideration (NSHC): The notice provided an opportunity to submit comments on the Commission's proposed NSHC determination. No comments have been received. The notice also provided an opportunity to request a hearing by June 13, 2010, which is within 60 days of the individual notice published on April 14; but indicated that if the Commission makes a final NSHC determination, any such hearing would take place after issuance of the amendment.
Date of initial individual notice in Federal Register: April 14, 2010 (75 FR 19428), followed by the repeat biweekly notice in the
Federal Register on May 4, 2010 (75 FR 23818).
The Commission's related evaluation of the amendment, state consultation, and final NSHC determination are contained in a Safety
Evaluation dated June 2, 2010.
Attorney for licensee: Mr. William Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Ave., White
Plains, NY 10601.
NRC Branch Chief: Robert J. Pascarelli.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: September 14, 2009, as supplemented on April 12, 2010.
Brief description of amendments: The amendments make miscellaneous administrative and editorial changes to the Technical Specifications
(TSs) and the Facility Operating Licenses (FOLs) including correction of typographical and format errors, correction of administrative differences between units, and deletion of historical requirements that have expired.
Date of issuance: June 15, 2010.
Effective date: As of the date of issuance, to be implemented within 60 days.
Amendment Nos.: 295 and 278.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments revised the TSs and the FOLs.
Date of initial notice in Federal Register: November 17, 2009 (74
FR 59262). The letter dated April 12, 2010, provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the application beyond the scope of the original Federal Register notice.
The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated June 15, 2010.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 18th day of June 2010.
For the Nuclear Regulatory Commission.
Robert A. Nelson,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
FR Doc. 2010-15439 Filed 6-28-10; 8:45 am
BILLING CODE 7590-01-P
