Federal Register: October 12, 2004 (Volume 69, Number 196)

DOCID: FR Doc 04-22544

SUBJECT CATEGORY:

Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

DOCUMENT SUMMARY:

I. Background

Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as amended (the Act), the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. The Act requires the Commission publish notice of any amendments issued, or proposed to be issued and grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued, from September 17, 2004, through September 30, 2004. The last biweekly notice was published on September 28, 2004 (69 FR 57978).
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Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene.

Normally, the Commission will not issue the amendment until the expiration of 60 days after the date of publication of this notice. The Commission may issue the license amendment before expiration of the 60 day period provided that its final determination is that the amendment involves no significant hazards consideration. In addition, the Commission may issue the amendment prior to the expiration of the 30 day comment period should circumstances change during the 30day comment period such that failure to act in a timely way would result, for example in derating or shutdown of the facility. Should the Commission take action prior to the expiration of either the comment period or the notice period, it will publish in the Federal Register a notice of issuance. Should the Commission make a final No Significant Hazards Consideration Determination, any hearing will take place after issuance. The Commission expects that the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 205550001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

Within 60 days after the date of publication of this notice, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's ``Rules of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested persons should consult a current copy of 10 CFR 2.309, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/ readingrm/doccollections/cfr/. If a request for a hearing or petition for leave to intervene is filed within 60 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.309, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) The name, address and telephone number of the requestor or petitioner; (2) the nature of the requestor's/petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the requestor's/petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order which may be entered in the proceeding on the requestor's/petitioner's interest. The petition must also set forth the specific contentions which the petitioner/requestor seeks to have litigated at the proceeding.

Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner/requestor shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner/ requestor intends to rely in proving the contention at the hearing. The petitioner/requestor must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner/requestor intends to rely to establish those facts or expert opinion. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner/requestor to relief. A petitioner/ requestor who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing.

If a hearing is requested, and the Commission has not made a final determination on the issue of no significant hazards consideration, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held. If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of
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the amendment. If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must be filed by: (1) First class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 205550001, Attention: Rulemaking and Adjudications Staff; (2) courier, express mail, and expedited delivery services: Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and Adjudications Staff; (3) Email addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, hearingdocket@nrc.gov; or (4) facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC, Attention: Rulemakings and Adjudications Staff at (301) 4151101, verification number is (301) 4151966. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 205550001, and it is requested that copies be transmitted either by means of facsimile transmission to (301) 4153725 or by email to
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

Nontimely requests and/or petitions and contentions will not be entertained absent a determination by the Commission or the presiding officer of the Atomic Safety and Licensing Board that the petition, request and/or the contentions should be granted based on a balancing of the factors specified in 10 CFR 2.309(a)(1)(i)(viii).

For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/readingrm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the NRC PDR Reference staff at 18003974209, 3014154737 or by email to
pdr@nrc.gov.
Entergy Nuclear Operations, Inc., Docket No. 50293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of amendment request: April 14, 2004.

Description of amendment request: The proposed amendment would make changes to the Technical Specifications (TSs) that will eliminate secondary containment operability requirements when handling sufficiently decayed irradiated fuel and performing core alterations, and will clarify requirements associated with operations with potential to drain the reactor vessel. This proposed amendment also uses Alternate Source Term (AST) methodology in accordance with 10 CFR 50.67 for calculating Fuel Handling Accident (FHA) consequences. The proposed amendment also removes TSs operability requirements for engineered safety features (ESF) (e.g. primary/secondary containment, standby gas treatment, and isolation capability) after the sufficient decay of ``recently'' irradiated fuel.

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. The Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The NRC staff's review is presented below.
1. Do the Proposed Changes Involve a Significant Increase in the Probability or Consequence of an Accident Previously Evaluated?

The proposed changes do not modify the design or operation of equipment used to handle and move new and spent fuel or to perform core alterations. The proposed amendment does not modify the design of the ESF equipment. The proposed changes, therefore, will not increase the probability of accidents previously evaluated.

AST analysis does not affect the performance of the systems or components used to mitigate the consequences of accidents previously evaluated. While a direct comparison between current methodologies used in the current Pilgrim design basis analysis and Regulatory Guide (RG) 1.183 is not possible due to different acceptance criteria, the AST calculations demonstrate that the radiological consequences to the accidents previously evaluated will still remain below the regulatory limits. Therefore, any potential change in the radiological consequences are not considered significant. Since the radiological consequences are below the regulatory limits and the probability of an accident is unchanged, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
2. Do the Proposed Changes Create the Possibility of a New or Different Kind of Accident From Any Accident Previously Analyzed?

There are no new plant operation modes or physical modifications being proposed. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously analyzed.
3. Do the Proposed Changes Involve a Significant Reduction in the Margin of Safety?

The licensee performed a comprehensive analysis and evaluation of the FHA using AST methodology and dose consequence analysis in accordance with 10 CFR 50.67. While direct comparison between methodologies used in the current Pilgrim design basis analysis and RG 1.183 is not possible due to different acceptance criteria, the revised doses will, however, remain below the total effective dose equivalent dose regulatory limits for the control room, exclusion area boundary, and low population zone as specified in 10 CFR 50.67. Therefore, by meeting the applicatory regulatory limits for AST, any potential decrease in a margin of safety would not be considered significant. The changes are, therefore, not considered 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 amendment request involves no significant hazards consideration.

Attorney for licensee: J. M. Fulton, Esquire, Assistant General Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, Massachusetts, 023605599.

NRC Acting Section Chief: Daniel S. Collins.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: June 17, 2004.

Description of amendment request: The proposed amendment would delete
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entries from Technical Specification (TS) Tables 3.2.6 and 4.2.6 related to the postaccident hydrogen and oxygen monitors. Licensees were generally required to implement upgrades as described in NUREG 0737, ``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory Guide (RG) 1.97, ``Instrumentation for LightWaterCooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.'' Implementation of these upgrades was an outcome of the lessons learned from the accident that occurred at TMI, Unit 2. Requirements related to combustible gas control were imposed by Order for many facilities and were added to or included in the TSs for nuclear power reactors currently licensed to operate. The revised 10 CFR 50.44, ``Combustible gas control system for nuclear power reactors,'' eliminated the requirements for hydrogen recombiners (not installed at Vermont Yankee and therefore not addressed by this proposed amendment) and relaxed safety
classifications and licensee commitments to certain design and qualification criteria for hydrogen and oxygen monitors.

The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on September 25, 2003 (68 FR 55416). The licensee affirmed the applicability of the model NSHC determination in its application dated June 17, 2004.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below: Criterion 1 The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The revised 10 CFR 50.44 no longer defines a designbasis loss ofcoolant accident (LOCA) hydrogen release, and eliminates requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a designbasis LOCA. The Commission has found that this hydrogen release is not risksignificant because the design basis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risksignificant accident sequences that could threaten containment integrity.

With the elimination of the designbasis LOCA hydrogen release, hydrogen and oxygen monitors are no longer required to mitigate designbasis accidents and, therefore, the hydrogen monitors do not meet the definition of a safetyrelated component as defined in 10 CFR 50.2. RG 1.97 Category 1, is intended for key variables that most directly indicate the accomplishment of a safety function for designbasis accident events. The hydrogen and oxygen monitors no longer meet the definition of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR 50.44, the Commission found that Category 3, as defined in RG 1.97, is an appropriate categorization for the hydrogen monitors because the monitors are required to diagnose the course of beyond designbasis accidents. Also, as part of the rulemaking to revise 10 CFR 50.44, the Commission found that Category 2, as defined in RG 1.97, is an appropriate categorization for the oxygen monitors, because the monitors are required to verify the status of the inert containment.

The regulatory requirements for the hydrogen and oxygen monitors can be relaxed without degrading the plant emergency response. The emergency response, in this sense, refers to the methodologies used in ascertaining the condition of the reactor core, mitigating the consequences of an accident, assessing and projecting offsite releases of radioactivity, and establishing protective action recommendations to be communicated to offsite authorities. Classification of the hydrogen monitors as Category 3,
classification of the oxygen monitors as Category 2, and removal of the hydrogen and oxygen monitors from TSs will not prevent an accident management strategy through the use of the severe accident management guidelines, the emergency plan, the emergency operating procedures, and site survey monitoring that support modification of emergency plan protective action recommendations.

Therefore, the relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TSs, does not involve a significant increase in the probability or the consequences of any accident previously evaluated.
Criterion 2 The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated

The relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TSs, will not result in any failure mode not previously analyzed. The hydrogen and oxygen monitor equipment was intended to mitigate a designbasis hydrogen release. The hydrogen and oxygen monitor equipment are not considered accident precursors, nor does their existence or elimination have any adverse impact on the preaccident state of the reactor core or post accident confinement of radionuclides within the containment building.

Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated. Criterion 3 The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety

The relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TSs, in light of existing plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, results in a neutral impact to the margin of safety.

The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a designbasis LOCA. The Commission has found that this hydrogen release is not risksignificant because the designbasis LOCA hydrogen release does not contribute to the conditional probability of a large release up to approximately 24 hours after the onset of core damage.

Category 3 hydrogen monitors are adequate to provide rapid assessment of current reactor core conditions and the direction of degradation while effectively responding to the event in order to mitigate the consequences of the accident. The intent of the requirements established as a result of the TMI, Unit 2 accident can be adequately met without reliance on safetyrelated hydrogen monitors.
Category 2 Oxygen Monitors Are Adequate To Verify the Status of an Inerted Containment.

Therefore, this change does not involve a significant reduction in the margin of safety. The intent of the requirements established as a result of the TMI, Unit 2 accident can be adequately met without reliance on safetyrelated oxygen monitors. Removal of hydrogen and oxygen monitoring from TSs will not result in a significant reduction in their functionality, reliability, and availability.

Based on the reasoning presented above and the previous discussion of the amendment request, the requested change does not involve a significant hazards consideration.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 200371128.

NRC Section Chief: Allen G. Howe.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: September 1, 2004.

Description of amendment request: The proposed amendment would delete Technical Specification (TS) 6.6.A, ``Occupational Radiation Exposure Report,'' and TS 6.6.B, ``Monthly Operating Reports.''

The NRC staff issued a notice of availability of a model no significant
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hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on June 23, 2004 (69 FR 35067). The licensee affirmed the applicability of the model NSHC determination in its application dated September 1, 2004.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration 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 eliminates the TS reporting requirements to provide a monthly operating report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the TS reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. 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 a physical alteration of the plant, add any new equipment, or require any existing equipment to be operated in a manner different from the present design. 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.

This is an administrative change to reporting requirements of plant operating information and occupational radiation exposure data, and has no effect on plant equipment, operating practices or safety analyses assumptions. For these reasons, the proposed change does not involve a significant reduction in the margin of safety.

Based upon the reasoning presented above, the requested change does not involve a significant hazards consideration.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 200371128.

NRC Section Chief: Allen G. Howe.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50 277 and 50278, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania

Date of amendment request: April 30, 2004.

Description of amendment request: The proposed change allows entry into a mode or other specified condition in the applicability of a technical specification (TS), while in a condition statement and the associated required actions of the TS, provided the licensee performs a risk assessment and manages risk consistent with the program in place for complying with the requirements of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.65(a)(4). Limiting Condition for Operation (LCO) 3.0.4 exceptions in individual TSs would be eliminated, several notes or specific exceptions are revised to reflect the related changes to LCO 3.0.4, and Surveillance Requirement (SR) 3.0.4 is revised to reflect the LCO 3.0.4 allowance.

This change was proposed by the industry's Technical Specification Task Force (TSTF) and is designated TSTF359. The NRC staff issued a notice of opportunity for comment in the Federal Register on August 2, 2002 (67 FR 50475), on possible amendments concerning TSTF359, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on April 4, 2003 (68 FR 16579). The licensee affirmed the applicability of the following NSHC determination in its application dated April 30, 2004.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below: Criterion 1The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. Being in a TS condition and the associated required actions is not an initiator of any accident previously evaluated. Therefore, the probability of an accident previously evaluated is not significantly increased. The consequences of an accident while relying on required actions as allowed by proposed LCO 3.0.4, are no different than the
consequences of an accident while entering and relying on the required actions while starting in a condition of applicability of the TS. Therefore, the consequences of an accident previously evaluated are not significantly affected by this change. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. Criterion 2The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From any Previously Evaluated

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed). Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the associated required actions of the TS, will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously evaluated. The addition of a requirement to assess and manage the risk introduced by this change will further minimize possible concerns. Thus, this change does not create the possibility of a new or different kind of accident from an accident previously evaluated.
Criterion 3The Proposed Change Does Not Involve a Significant Reduction in a Margin of Safety

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. The TS allow operation of the plant without the full complement of equipment through the conditions for not meeting the TS LCO. The risk associated with this allowance is managed by the imposition of required actions that must be performed within the prescribed completion times. The net effect of being in a TS condition on the margin of safety is not considered significant. The proposed change does not alter the required actions or completion times of the TS. The proposed change allows TS conditions to be entered, and the associated required actions and completion times to be used in new circumstances. This use is predicated upon the licensee's
performance of a risk assessment and the management of plant risk. The change also eliminates current allowances for utilizing required actions and completion times in similar circumstances, without assessing and managing risk. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

The NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for Licensee: Thomas S. O'Neill, Associate and General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Section Chief: Daniel Collins, Acting.
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FirstEnergy Nuclear Operating Company, Docket No. 50346, DavisBesse Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: August 2, 2004.

Description of amendment request: The proposed amendment would delete Technical Specification (TS) Section 6.8.4, ``Post Accident Sampling,'' and the related requirements to maintain a PostAccident Sampling System (PASS). Licensees were generally required to implement PASS upgrades as described in NUREG0737, ``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97, Revision 3, ``Instrumentation for LightWaterCooled Nuclear Power Plants to Access Plant and Environs Conditions During and Following an Accident.'' Implementation of these upgrades was an outcome of the U.S. Nuclear Regulatory Commission's (NRC) lessons learned from the accident that occurred at TMI Unit 2. Requirements related to PASS were imposed by Order for many facilities and were added to or included in the TS for nuclear power reactors currently licensed to operate. Lessons learned and improvements implemented over the last 20 years have shown that the information obtained from PASS can be readily obtained through other means or is of little use in the assessment and mitigation of accident conditions.

The NRC staff issued a notice of opportunity for comment in the Federal Register on March 3, 2003 (68 FR 10052) on possible amendments to eliminate PASS, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated line item improvement process. The NRC staff subsequently issued a notice of availability of the models for referencing in a license amendment application in the Federal Register on May 13, 2003 (68 FR 25664). The licensee affirmed the applicability of the following NSHC determination in its application dated August 2, 2004.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below: Criterion 1The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The PASS was originally designed to perform many sampling and analysis functions. These functions were designed and intended to be used in post accident situations and were put into place as a result of the TMI2 accident. The specific intent of the PASS was to provide a system that has the capability to obtain and analyze samples of plant fluids containing potentially high levels of radioactivity, without exceeding plant personnel radiation exposure limits. Analytical results of these samples would be used largely for verification purposes in aiding the plant staff in assessing the extent of core damage and subsequent offsite radiological dose projections. The system was not intended to and does not serve a function for preventing accidents and its elimination would not affect the probability of accidents previously evaluated.

In the 20 years since the TMI2 accident and the consequential promulgation of post accident sampling requirements, operating experience has demonstrated that a PASS provides little actual benefit to post accident mitigation. Past experience has indicated that there exists inplant instrumentation and methodologies available in lieu of a PASS for collecting and assimilating information needed to assess core damage following an accident. Furthermore, the implementation of Severe Accident Management Guidance (SAMG) emphasizes accident management strategies based on inplant instruments. These strategies provide guidance to the plant staff for mitigation and recovery from a severe accident. Based on current severe accident management strategies and guidelines, it is determined that the PASS provides little benefit to the plant staff in coping with an accident.

The regulatory requirements for the PASS can be eliminated without degrading the plant emergency response. The emergency response, in this sense, refers to the methodologies used in ascertaining the condition of the reactor core, mitigating the consequences of an accident, assessing and projecting offsite releases of radioactivity, and establishing protective action recommendations to be communicated to offsite authorities. The elimination of the PASS will not prevent an accident management strategy that meets the initial intent of the postTMI2 accident guidance through the use of the SAMGs, the emergency plan (EP), the emergency operating procedures (EOP), and site survey monitoring that support modification of emergency plan protective action recommendations (PARs).

Therefore, the elimination of PASS requirements from Technical Specification (TS) (and other elements of the licensing bases) does not involve a significant increase in the consequences of any accident previously evaluated.
Criterion 2The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From any Previously Evaluated

The elimination of PASS related requirements will not result in any failure mode not previously analyzed. The PASS was intended to allow for verification of the extent of reactor core damage and also to provide an input to offsite dose projection calculations. The PASS is not considered an accident precursor, nor does its existence or elimination have any adverse impact on the preaccident state of the reactor core or post accident confinement of radioisotopes within the containment building.

Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated. Criterion 3The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety

The elimination of the PASS, in light of existing plant equipment, instrumentation, procedures, and programs that provide effective mitigation of and recovery from reactor accidents, results in a neutral impact to the margin of safety. Methodologies that are not reliant on PASS are designed to provide rapid assessment of current reactor core conditions and the direction of degradation while effectively responding to the event in order to mitigate the consequences of the accident. The use of a PASS is redundant and does not provide quick recognition of core events or rapid response to events in progress. The intent of the requirements established as a result of the TMI2 accident can be adequately met without reliance on a PASS.

Therefore, this change does not involve a significant reduction in the margin of safety.

Based upon the reasoning presented above and the previous discussion of the amendment request, the requested change does not involve a significant hazards consideration.

The NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Section Chief: Anthony J. Mendiola.
Indiana Michigan Power Company, Docket Nos. 50315 and 50316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of amendment requests: April 13, 2004.

Description of amendment requests: The proposed amendments would change the licensing basis as described in the Updated Final Safety Analysis Report to allow the use of a reinforcing bar (rebar) yield strength value based on measured material properties, as documented in the licensee rebar acceptance tests, in control rod drive missile shield structural calculations.

1. Does the proposed change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?
[[Page 60683]]

Response: No
Probability of Occurrence of an Accident Previously Evaluated

This is a change in the method of determining the acceptability of accommodating the pressure load following a lossofcoolant accident. No physical changes are being made to the plant and no potential accident initiators are introduced by this change. Thus, the probability of the occurrence of any accident previously evaluated is not significantly increased.

Consequences of an Accident Previously Evaluated

There is reasonable assurance that the ability of control rod drive missile shields (missile shields) to maintain their structural capability and continue to function as a part of the divider barrier separating the lower containment from the upper containment is not impacted by this change. The data obtained from rebar acceptance test reports demonstrate that the missile shields have adequate strength to accommodate the load that would be imposed under assumed accident conditions. As a result, the consequences of any accident previously evaluated are not significantly 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 use of increased missile shield rebar yield strength for the missile shield structural capability under accident conditions does not alter the evaluation of the missile shields' structural capability during normal operation, the operational condition in which a new or different kind of accident would be initiated. The change does not physically alter plant components nor does it alter plant operation. The change does not adversely affect current system interfaces or create new interfaces that could result in an accident or malfunction of a different kind than previously evaluated.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No

The margin of safety for the missile shields is provided by the factors that are applied to the individual loads determining the load imposed on the missile shields under accident conditions. These code safety factors are sufficient to ensure that both anticipated and unanticipated loads can be withstood by the concrete structures. The use of yield strengths based on measured material properties as documented in the I&M [Indiana Michigan Power Company] rebar acceptance tests for the missile shield structural evaluation has no effect on the margin of safety provided by the load safety factors. I&M continues to use the same load factors that were used to license the Donald C. Cook Nuclear Plant.

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 requests involve no significant hazards consideration.

Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, Buchanan, MI 49107.

NRC Section Chief: L. Raghavan.
Omaha Public Power District, Docket No. 50285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: September 7, 2004.

Description of amendment request: The proposed amendment revises Fort Calhoun Station (FCS) Technical Specification (TS) 5.9.5, ``Core Operating Limits Report,'' such that it will read consistent with TS 5.6.5 of NUREG1432, Standard Technical SpecificationsCombustion Engineering Plants. In addition, the list of core reload analysis methodologies contained in TS 5.9.5b used to determine the core operating limits is updated to move many of these references to Omaha Public Power District (OPPD) core reload analysis methodology documents OPPDNA8301, 8302, and 8303. Several analytical method references that are no longer applicable to FCS are deleted from TS 5.9.5b; several references will remain, as they are not suitable for incorporation into the core reload analysis documents.

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes are primarily administrative in nature to achieve consistency with Standard Technical Specifications and to update the list of NRC reviewed and approved analytical methods used to develop core operating limits. Several of the analytical methods are no longer applicable to Fort Calhoun Station, Unit No. 1, (FCS) and thus are deleted from the Technical Specifications (TSs). Many of the topical reports currently referenced in TS 5.9.5b are more suitably referenced in the OPPD core reload methodology documents where they have been relocated.

The OPPD core reload methodology documents remain referenced in TS 5.9.5b and as such are subject to NRC review and approval. The relocation of the topical reports referenced in TS 5.9.5b to OPPD core reload methodology documents is an administrative change. In addition to the incorporation of references currently found in TS 5.9.5b, OPPD core reload methodology documents OPPDNA8301, 8302, and 8303 are revised to remove characters designating them as proprietary, and approved. This is an administrative change, as OPPD no longer considers the documents to be proprietary or topical reports. OPPD core reload methodology documents OPPDNA8301, 8302, and 8303 are enclosed for NRC review and approval [attached to the licensee's September 7, 2004, letter] of the changes noted above and incorporation of the CASMO4 (C4) computer code, which is described below.

OPPD is adding the C4 code to OPPDNA8302, Reload Core Analysis Methodology, Neutronics Design Methods and Verification and will use the code for nuclear design analysis. This will allow the use of the C4 and SIMULATE3 (S3) methodology to perform all steadystate pressurized water reactor (PWR) core physics analyses. The probability of occurrence of an accident previously evaluated will not be increased by the proposed change in the particular codes used for physics calculations for nuclear design analysis. The results of nuclear design analyses are used as inputs to the analysis of accidents that are evaluated in the Updated Safety Analysis Report (USAR). These inputs do not alter the physical characteristics or modes of operation of any system, structure, or component involved in the initiation of an accident. Thus, there is no significant increase in the probability of an accident previously evaluated as a result of this change.

The consequences of an accident evaluated in the USAR are affected by the value of inputs to the transient safety analysis. An extensive benchmark of C4/S3 predictions was performed with measured data using a variety of fuel designs and operating conditions in power reactors and critical experiments. The accuracy of C4/S3 is similar to, and sometimes better than, the accuracy of C3/S3. Furthermore, there is always the potential for the value of the nuclear design parameters to change solely as a result of the new core reload fuel core loading pattern. Regardless of the source of a change, an assessment is always made of changes to the nuclear design parameters with respect to their effects on the consequences of accidents previously evaluated in the USAR. Refueling is an anticipated activity, which is described in the USAR. If increased consequences are anticipated, compensatory actions are implemented to neutralize any expected increase in consequences. These compensatory actions include, but are not limited to, crediting any existing margins in the analysis or redefining the operating envelope to avoid increased consequences. Thus, the nuclear design parameters are intermediate results and by themselves will not result in an increase in the consequence of an accident evaluated in the USAR.

[[Page 60684]]

Therefore, the use of the C4/S3 code package, which will perform the same functions as the C3/S3 codes with similar accuracy, does not significantly increase the consequences of an accident previously evaluated.

2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes are primarily administrative in nature. The changes achieve consistency with Standard Technical Specifications, update the list of NRC reviewed and approved analytical methods used to develop core operating limits by deleting certain analytical methods no longer applicable to FCS and relocating many of the remainder to OPPD core reload analysis methodology documents, and make minor administrative changes to OPPD core reload analysis documents referenced in TS 5.9.5b. OPPD intends to utilize the C4/ S3 code package for nuclear design analysis. The proposed amendment would add the C4 code to OPPD core reload analysis methodology document OPPDNA8302.

The possibility for a new or different kind of accident evaluated previously in the USAR will not be created by the proposed administrative changes or the change to the particular codes used for physics calculations for nuclear design analyses. The change involves adding the Studsvik C4 code to OPPD core reload analysis methodology document OPPDNA8302. The C4 code is an update to the C3 code currently approved for use at FCS. The results of nuclear design analyses are used as inputs to the analysis of accidents that are evaluated in the USAR. These inputs do not alter the physical characteristics or modes of operation of any system, structure or component involved in the initiation of an accident. Therefore, these administrative changes and the addition of the C4 code, which will perform the same functions, as the C3 code with similar accuracy, does not increase the possibility of a new or different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction in a margin of safety.

The proposed change does not involve a significant reduction in a margin of safety. The margin of safety as defined in the basis for any technical specification will not be reduced nor increased by the proposed administrative changes or the change to the codes used for physics calculations for nuclear design analyses. The changes achieve consistency with Standard Technical Specifications, update the list of NRC approved analytical methods used to develop core operating limits by deleting certain analytical methods no longer applicable to FCS and relocating many of the remainder to OPPD core reload analysis methodology documents, and make minor administrative changes to OPPD core reload analysis documents referenced in TS 5.9.5b.

The change involves the addition of the Studsvik C4 code to OPPD core reload analysis methodologies for nuclear design analysis. Extensive benchmarking of the C4/S3 computer codes has
demonstrated that the values of those parameters used in the safety analysis are not significantly changed relative to the values obtained using the NRC approved C3/S3 computer codes. For any changes in the calculated values that do occur, the application of appropriate biases and uncertainties ensures that the current margin of safety is maintained. Specifically, use of these code specific biases and uncertainties in safety evaluations continues to provide the same statistical assurance that the values of the nuclear parameters used in the safety analysis are conservative with respect to the actual values on at least a 95/95 probability/confidence basis.

Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC 200053502.

NRC Section Chief: Robert Gramm.
PSEG Nuclear LLC, Docket Nos. 50272 and 50311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

Date of amendment request: April 15, 2004, as supplemented August 11, 2004.

Description of amendment request: The amendment request proposes to revise the Salem Unit No. 1 Technical Specifications (TSs) to reflect the addition of the chilled water system to provide cooling water to the containment fan cooling units (CFCUs). The amendment request also proposes to revise a nonconservative Action Statement for Salem Unit Nos. 1 and 2 that allows three containment cooling fans to be inoperable under certain conditions.

1. Does the change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?

Response: No.

Containment cooling fans remove containment heat loads under both normal and accident conditions. As such, they have no impact on the probability of occurrence of any previously evaluated accidents, although they do function to mitigate accident consequences. With regard to accident consequences, revised containment response analysis has been performed with the proposed changes of this license amendment. This analysis demonstrates that containment pressure and temperature limits continue to be met as further described below.

The addition of the nonsafety related chilled water system does not represent an increase in the consequences of an accident since, at the onset of the accident, the chilled water supply is
automatically isolated on the resulting safety injection signal and the safety related Service Water System supplies the cooling method to remove the containment heat loads, as presently analyzed. Analysis has been performed to evaluate any potential failures that could prevent the Containment Cooling System to perform [sic] its safety related functions. Redundancy in the chilled water system and transfer to service water during an accident are incorporated in the design. In addition, as a conservative measure, an action statement has been added to require prompt action to restore containment cooling or commence a unit shutdown in the event of an unexpected condition that results in the loss of normal containment cooling capability.

The accidents previously evaluated that are associated with containment heat removal are design basis lossofcoolant accident (LOCA) and main steam line break (MSLB) accident. In the case of the design basis LOCA, the revised analysis demonstrates that all cases resulted in a peak containment pressure that was less than 47 psig. In addition, all longterm cases were well below 50% of the peak value within 24 hours. Based on the results, applicable criteria for Salem Unit 1 have been met and therefore, the consequences of previously evaluated accidents are not increased.

The proposed change to the nonconservative TS 3.6.2.3 Action b, maintains that five CFCUs remain operable to ensure that, upon a single failure, a minimum of three CFCUs will provide the required containment and air mixing which is consistent with the current Salem Dose Analysis.

Consequently, the proposed license amendment does not increase the probability of occurrence or the consequences of accidents previously evaluated for Salem.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

Response: No.

Containment cooling fans remove containment heat loads under both normal and accident conditions. The containment cooling fans are presently part of the plant protection equipment and have been analyzed and evaluated as to their function and effectiveness. Consequently, they cannot create the possibility of any new or different kinds of accidents from any previously evaluated. The addition of a chilled water system that is isolated on an accident condition does not create a new or different kind of accident. The accidents analyzed are the LOCA and MSLB, which are part of the Salem Design Bases.

The proposed change to the nonconservative TS 3.6.2.3 Action b, maintains that five CFCUs remain operable to ensure that, upon a single failure, a minimum of
[[Page 60685]]
three CFCUs will provide the required containment and air mixing which is consistent with the current Salem Dose Analysis.

Therefore, the proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the change involve a significant reduction in a margin of safety?

Response: No.

The margin of safety pertinent to the proposed changes is the dose consequences resulting from a design basis LOCA. Containment cooling fans affect potential dose consequences in that they assist in maintaining containment pressure and temperature within design limits. By maintaining these limits, three critical functions are performed. These are:

a. Containment integrity is assured by maintaining pressure below the containment design limit.

b. By maintaining pressure below 47 psig, leakage of containment atmosphere to the surrounding environment is retained within the leakage testing results of 10 CFR 50, Appendix J. In this case, the Appendix J testing procedures provide the margin of safety, as long as the limiting pressure (47 psig) is not exceeded.

c. By maintaining containment temperature within limits, the qualification of vital electrical equipment to function in the post accident containment environment is assured. In this case, the margin of safety is provided by the testing and evaluation procedures implemented by 10 CFR 50.49.

In addition, as a conservative measure, an action statement has been added to require prompt action to restore containment cooling or commence a unit shutdown in the event of an unexpected condition that results in the loss of normal containment cooling capability.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business UnitN21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Section Chief: James W. Clifford.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50244, R. E. Ginna Nuclear Power Plant, Wayne County, New York

Date of amendment request: July 26, 2004.

Description of amendment request: The proposed amendment would delete Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure Report,'' and TS 5.6.4, ``Monthly Operating Reports.''

The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on June 23, 2004 (69 FR 35067). The licensee affirmed the applicability of the model NSHC determination in its application dated July 26, 2004.

The proposed change eliminates the TS reporting requirements to provide a monthly operating report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the Technical Specification reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Criterion 2 The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated?

Based upon the reasoning presented above, the NRC staff proposes to determine that the requested change does not involve significance hazards consideration.

Attorney for licensee: Daniel F. Stenger, Ballard Spahr Andrews & Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 20005.

NRC Section Chief: Richard J. Laufer.
South Carolina Electric & Gas Company, South Carolina Public Service Authority, Docket No. 50395, Virgil C. Summer Nuclear Station (VCSNS), Unit No. 1, Fairfield County, South Carolina

Date of amendment request: August 24, 2004.

Description of amendment request: The proposed change will revise Surveillance Requirements (SRs) 4.7. 1.2.a.1 and 4.7 .1.2.a.2 to reflect a more representative model of the Emergency Feedwater (EFW) System. The new model has established new technical specification (TS) acceptance criteria to assure the design requirements of the system are met. These required characteristics are more stringent than those currently in the VCSNS TSs for this system.

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

This change represents a more restrictive surveillance requirement than currently exists for TS Surveillance 4.7. 1.2.a.1 and 4.7 .1.2.a.2. These proposed surveillance acceptance criteria changes will ensure that the motor driven EFW pumps and the turbine driven EFW pump can continue to perform their design function. There are no changes planned to any plant installed hardware or software and normal plant operations will not be impacted.

The probability or consequences of accidents previously evaluated in the VCSNS FSAR [Final Safety Analysis Report] are un

SUMMARY:

Operating licenses, amendments; no significant hazards considerations; biweekly notices,

DOCUMENT BODY 2:

I. Background

Date of amendment request: April 14, 2004.

Attorney for licensee: J. M. Fulton, Esquire, Assistant General Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, Massachusetts, 023605599.

NRC Acting Section Chief: Daniel S. Collins.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: June 17, 2004.

Based on the reasoning presented above and the previous discussion of the amendment request, the requested change does not involve a significant hazards consideration.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 200371128.

NRC Section Chief: Allen G. Howe.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: September 1, 2004.

Description of amendment request: The proposed amendment would delete Technical Specification (TS) 6.6.A, ``Occupational Radiation Exposure Report,'' and TS 6.6.B, ``Monthly Operating Reports.''

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

Based upon the reasoning presented above, the requested change does not involve a significant hazards consideration.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 200371128.

Date of amendment request: April 30, 2004.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below: Criterion 1The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The proposed change allows entry into a mode or other specified condition in the applicability of a TS, while in a TS condition statement and the associated required actions of the TS. The TS allow operation of the plant without the full complement of equipment through the conditions for not meeting the TS LCO. The risk associated with this allowance is managed by the imposition of required actions that must be performed within the prescribed completion times. The net effect of being in a TS condition on the margin of safety is not considered significant. The proposed change does not alter the required actions or completion times of the TS. The proposed change allows TS conditions to be entered, and the associated required actions and completion times to be used in new circumstances. This use is predicated upon the licensee's
performance of a risk assessment and the management of plant risk. The change also eliminates current allowances for utilizing required actions and completion times in similar circumstances, without assessing and managing risk. The net change to the margin of safety is insignificant. Therefore, this change does not involve a significant reduction in a margin of safety.

The NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for Licensee: Thomas S. O'Neill, Associate and General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Section Chief: Daniel Collins, Acting.
[[Page 60682]]
FirstEnergy Nuclear Operating Company, Docket No. 50346, DavisBesse Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: August 2, 2004.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below: Criterion 1The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated. Criterion 3The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety

Therefore, this change does not involve a significant reduction in the margin of safety.

Based upon the reasoning presented above and the previous discussion of the amendment request, the requested change does not involve a significant hazards consideration.

The NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Section Chief: Anthony J. Mendiola.
Indiana Michigan Power Company, Docket Nos. 50315 and 50316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of amendment requests: April 13, 2004.

1. Does the proposed change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?
[[Page 60683]]

Response: No
Probability of Occurrence of an Accident Previously Evaluated

Consequences of an Accident Previously Evaluated

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No

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

Therefore, the proposed change does not involve a significant reduction in the margin of safety.

Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, Buchanan, MI 49107.

NRC Section Chief: L. Raghavan.
Omaha Public Power District, Docket No. 50285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: September 7, 2004.

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

[[Page 60684]]

2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction in a margin of safety.

Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC 200053502.

NRC Section Chief: Robert Gramm.
PSEG Nuclear LLC, Docket Nos. 50272 and 50311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

Date of amendment request: April 15, 2004, as supplemented August 11, 2004.

1. Does the change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?

Response: No.

Consequently, the proposed license amendment does not increase the probability of occurrence or the consequences of accidents previously evaluated for Salem.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

Response: No.

Therefore, the proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the change involve a significant reduction in a margin of safety?

Response: No.

a. Containment integrity is assured by maintaining pressure below the containment design limit.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business UnitN21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Section Chief: James W. Clifford.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50244, R. E. Ginna Nuclear Power Plant, Wayne County, New York

Date of amendment request: July 26, 2004.

Description of amendment request: The proposed amendment would delete Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure Report,'' and TS 5.6.4, ``Monthly Operating Reports.''

The NRC staff issued a notice of availability of a model no significant hazards consideration (NSHC) determination for referencing in license amendment applications in the Federal Register on June 23, 2004 (69 FR 35067). The licensee affirmed the applicability of the model NSHC determination in its application dated July 26, 2004.

The proposed change eliminates the TS reporting requirements to provide a monthly operating report of shutdown experience and operating statistics if the equivalent data is submitted using an industry electronic database. It also eliminates the Technical Specification reporting requirement for an annual occupational radiation exposure report, which provides information beyond that specified in NRC regulations. The proposed change involves no changes to plant systems or accident analyses. As such, the change is administrative in nature and does not affect initiators of analyzed events or assumed mitigation of accidents or transients. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Criterion 2 The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Accident Previously Evaluated?

Based upon the reasoning presented above, the NRC staff proposes to determine that the requested change does not involve significance hazards consideration.

Attorney for licensee: Daniel F. Stenger, Ballard Spahr Andrews & Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC 20005.

Date of amendment request: August 24, 2004.

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

The probability or consequences of accidents previously evaluated in the VCSNS FSAR [Final Safety Analysis Report] are un

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Federal Register: October 12, 2004 (Volume 69, Number 196)

DOCID: FR Doc 04-22544

NUCLEAR REGULATORY COMMISSION

Nuclear Regulatory Commission

NOTICE: NOTICES

ACTION: Environmental statements; availability, etc.:

SUBJECT CATEGORY:

DOCUMENT SUMMARY:

Consequences of an Accident Previously Evaluated

[[Page 60684]]

SUMMARY:

DOCUMENT BODY 2:

Consequences of an Accident Previously Evaluated

[[Page 60684]]