Federal Register: August 31, 2004 (Volume 69, Number 168)

DOCID: FR Doc 04-19586

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, August 6 through August 19, 2004. The last biweekly notice was published on August 19, 2004 (69 FR 51487). 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
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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 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 3014153725 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.
AmerGen Energy Company, LLC, Docket No. 50461, Clinton Power Station, Unit 1, DeWitt County, Illinois

Date of amendment request: June 22, 2004.

Description of amendment request: The proposed amendment would revise Technical Specification 3.1.8, ``Scram Discharge Volume (SDV) Vent and Drain Valves,'' to allow a vent or drain line with one inoperable valve to be isolated instead of requiring the valve to be restored to Operable status within 7 days.

The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice of opportunity for comment in the Federal Register on February 24, 2003 (68 FR 8637), on possible amendments to revise the action for one or more SDV vent or drain lines with an inoperable valve, including a model safety evaluation and model no significant hazards consideration (NSHC) determination, using the consolidated lineitem 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 15, 2003 (68 FR 18294). The licensee affirmed the applicability of the model NSHC determination in its application dated June 22, 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

A change is proposed to allow the affected SDV vent and drain line to be isolated when there are one or more SDV vent or drain lines with one valve inoperable instead of requiring the valve to be restored to operable status within 7 days. With one SDV vent or drain valve inoperable in one or more lines, the isolation function would be maintained since the redundant valve in the affected line would perform its safety function of isolating the SDV. Following the completion of the required action, the isolation function is fulfilled since the associated line is isolated. The ability to vent and drain the SDV is maintained and controlled through
administrative controls. This requirement assures the reactor protection system is not adversely affected by the inoperable valves. With the safety functions of the valves being maintained, the probability or consequences of an accident previously evaluated are not significantly increased.
Criterion 2The 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 involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. Thus, 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 proposed change ensures that the safety functions of the SDV vent and drain valves are fulfilled. The isolation function is maintained by redundant valves and by the required action to isolate the affected line. The ability to vent and drain the SDV is maintained through administrative controls. In addition, the reactor protection system will prevent filling of the SDV to the point that it has insufficient volume to accept a full scram. Maintaining the safety functions related to isolation of the SDV and insertion of control rods ensures that the proposed change does not involve a significant reduction in the margin of safety.

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

Attorney for licensee: Mr. Thomas S. O'Neill, Associate General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 60666.

NRC Section Chief: Anthony J. Mendiola.
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AmerGen Energy Company, LLC, Docket No. 50289, Three Mile Island Nuclear Station, Unit 1 (TMI1), Dauphin County, Pennsylvania

Date of amendment request: April 23, 2004.

Description of amendment request: The proposed amendment would delete Technical Specification (TS) Section 6.16, ``PostAccident Sampling Programs NUREG 0737 (II.B.3, IIF.1.2),'' 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 NRC's 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 TSs 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 April 23, 2004.

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 Specifications (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: Thomas S. O'Neill, Associate General Counsel, AmerGen Energy Company, LLC, 4300 Winfield Road, Warrenville, IL 60555.

NRC Section Chief: Richard J. Laufer.
Carolina Power & Light Company, Docket Nos. 50325 and 50324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina

Date of amendments request: July 26, 2004.

Description of amendments request: The proposed amendments would delete requirements from the Technical Specifications (TS) to maintain hydrogen recombiners and hydrogen and oxygen monitors. Licensees were generally required to implement upgrades as described in NUREG0737, ``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory Guide (RG) 1.97, ``Instrumentation for LightWater Cooled 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
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many facilities and were added to or included in the TS for nuclear power reactors currently licensed to operate. The revised 10 CFR 50.44, ``Combustible gas control for nuclear power reactors,'' eliminated the requirements for hydrogen recombiners 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 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 no significant hazards consideration determination in its application dated July 26, 2004.

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 TS will not prevent an accident management strategy through the use of the SAMGs [severe accident management guidelines], the emergency plan (EP), the emergency operating procedures (EOPs), and site survey monitoring that support modification of emergency plan protective action recommendations (PARs).

Therefore, the elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, does not involve a significant increase in the probability or 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 the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, will not result in any failure mode not previously analyzed. The hydrogen recombiner and hydrogen and oxygen monitor equipment was intended to mitigate a designbasis hydrogen release. The hydrogen recombiner and 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.

The elimination of the hydrogen recombiner requirements and relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, 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 TS will not result in a significant reduction in their functionality, reliability, and availability.

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

Attorney for licensee: Steven R. Carr, Associate General Counsel Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Section Chief (Acting): Michael L. Marshall. Carolina Power & Light Company, Docket No. 50261, H. B. Robinson Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: June 21, 2004.

Description of amendment request: The proposed amendment would revise Technical Specification Section 5.5.14, ``Technical Specifications (TS) Bases Control Program,'' to replace the previous 10 CFR 50.59 term ``unreviewed safety question'' with current terminology. The proposed amendment would also revise TS Section 5.7.1, ``High Radiation Area,'' to add wording that was inadvertently deleted with the issuance of the Improved Standard Technical Specifications in Amendment No. 176.

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

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The proposed changes do not modify the facility or the procedures for operation of the facility. One change updates the terminology used in 10 CFR 50.59 evaluations. The change does not alter the requirement of the TS Bases Control Program. The requirement for NRC review and approval of a TS Bases change is still determined through the use of the 10 CFR 50.59 review process. The second change corrects a typographical error that occurred under Amendment No. 176. The wording as proposed in this correction restores the requirement to the phraseology approved in Amendment No. 152 and is consistent with existing plant procedures.

Since there are no changes to the facility or facility procedures, the proposed changes do not involve a significant increase in the probability or 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 previously evaluated.

Since there are no changes to the facility or facility procedures, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

The proposed changes continue to provide the controls necessary to ensure changes to the TS Bases are made in conformance with 10 CFR 50.59. The proposed changes continue to provide the controls necessary to ensure adequate control of High Radiation Areas. The proposed changes will not result in any changes to the facility or facility operating procedures. Therefore, the changes do not result in a significant reduction in the margin of safety.

Based on the above discussion, Carolina Power & Light has determined that the requested change does not involve a significant hazards consideration.

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: Steven R. Carr, Associate General Counsel Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Section Chief: Michael L. Marshall, Acting.
Energy Northwest, Docket No. 50397, Columbia Generating Station, Benton County, Washington

Date of amendment request: June 9, 2004.

Description of amendment request: The proposed change revises Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.4.11, ``RCS Pressure and Temperature (P/T) Limits,'' to replace the P/T curves for inservice leak and hydrostatic testing, nonnuclear heating and cooldown, and nuclear heating and cooldown currently illustrated in TS Figures 3.4.111, 3.4.112, and 3.4.113,

respectively.

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

The proposed changes deal exclusively with the Reactor Coolant System (RCS) Pressure and Temperature (P/T) curves, which define the limitations for operation and testing. Because of the design conservatisms used to calculate the RCS P/T limits, reactor vessel failure has a low probability of occurrence and is not considered as a design basis accident in the safety analyses of the plant. The proposed changes adjust the reference temperature for the limiting material to account for irradiation effects and provide a comparable level of protection as previously evaluated and approved. The adjusted reference temperature calculations were performed in accordance with the requirements of 10 CFR [Part] 50 Appendix G using the guidance contained in RG [Regulatory Guide] 1.99, Revision 2, ``Radiation Embrittlement of Reactor Vessel Materials,'' to provide operating limits for up to 33.1 EFPY [effective full power years]. The proposed license amendment does not involve a change to operation of equipment required to mitigate any accident analyzed in Columbia's UFSAR [Updated Final Safety Analysis Report]. Therefore, the proposed change does not involve a significant increase in the probability or 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 revised P/T curves are based on a later edition and addenda of the ASME Code that incorporates current industry standards for the curves. The revised curves are also based on an RPV [reactor pressure vessel] fluence that has been recalculated in accordance with the methodology of RG 1.190. The proposed changes do not involve a modification to plant equipment. There is no effect on the function of any plant system, and no new system interactions are introduced by this change. No new failure modes are introduced. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

The proposed curves conform to the guidance contained in RG 1.190, ``Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,'' and RG 1.99, Revision 2, ``Radiation Embrittlement of Reactor Vessel Materials,'' and maintain the safety margins specified in 10 CFR [Part] 50 Appendix G. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Attorney for licensee: Thomas C. Poindexter, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC 200053502.

NRC Section Chief: Stephen Dembek.
Energy Northwest, Docket No. 50397, Columbia Generating Station, Benton County, Washington

Date of amendment request: August 5, 2004.

Description of amendment request: The proposed change will revise Technical Specification (TS) 5.5.12, ``Primary Containment Leakage Rate Testing Program,'' to allow a onetime deferral of the Type A containment integrated leak rate test (ILRT). The current 10year interval between Type A tests would be extended to 15 years from the previous time a Type A test was performed. The last Type A test was performed on July 20, 1994.

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

The proposed onetime extension to the Type A testing interval from onceper10 years to onceper15 years will not increase the probability of an accident previously evaluated. The performance of Type A tests is not an accident initiator. The primary containment Type A testing interval extension does not involve a plant
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modification and will not cause equipment failure or accident initiation.

The proposed extension to the Type A testing interval does not involve a significant increase in the consequences of an accident. The NUREG 1493 generic study of the effects of extending containment leakage testing concluded that Type B and C testing can identify the vast majority (greater than 95 percent) of potential leakage paths and that reducing the Type A test interval to onceper20 years leads to an ``imperceptible increase in risk.'' Other testing and inspection programs, in addition to the Type A test, provide a high degree of assurance that the primary containment integrity will be maintained. Inspections required by the Maintenance Rule and ASME Code [are] periodically performed in order to identify indications of containment degradation that could affect containment leak tightness.

Experience at Columbia demonstrates that excessive containment leakage paths are detectable by Type B and C local leak rate tests. Type B and C testing will identify containment openings, such as a valve, that would otherwise be detected by the Type A test. These factors show that a onetime Type A test interval extension from onceper10 years to onceper15 years will not involve a

significant increase in the consequences of an accident.

Previous Type A test results at Columbia show leakage has not exceeded acceptance criteria in the past, indicating a leaktight containment and demonstrating the structural capability of the primary containment. The testing results have established that Columbia has had acceptable containment leakage rates with considerable margin.

Therefore, the proposed changes do not involve a significant increase in the probability or 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 Columbia primary containment is designed to contain energy and fission products during and after a design basis accident. The proposed extension of the Type A testing interval will not create the possibility of a new or different type of accident from any previously evaluated. There are no changes being made to the physical plant or in operation of the plant that could introduce a new failure mode with the potential to create an accident or affect mitigation of an accident.

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

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

The proposed extension of the Type A testing interval will not significantly reduce the margin of safety. The NUREG 1493 generic study of the effects of extending containment leakage testing found that a 20year interval in Type A leakage testing leads to an ``imperceptible increase in risk.'' NUREG 1493 found that
generically, the design containment leakage rate contributes less than 0.1 percent to the overall accident risk and that the increase in the Type A testing interval would have a minimal effect on risk because the vast majority (greater than 95 percent) of all potential leakage paths are detected by Type B and C leakage testing.

A Columbia plant specific probabilistic risk assessment on the change in the Type A test interval from onceper10 years to once per15 years determined:

The risk impact due to a change in Large Early Release Frequency (LERF) is an increase of 2E8/year that is characterized by Regulatory Guide 1.174 [``An Approach for Using Probabilistic Risk Assessment in RiskInformed Decisions on PlantSpecific Changes to the Licensing Basis''] as ``very small.''

The total integrated plant risk increase measured by personrem/year is negligible.

The change in conditional containment failure probability is an increase of 0.1 percent, which is considered to represent a very small impact on risk.

Deferral of Type A testing for Columbia does not increase the level of risk to the public due to loss of capability to detect and measure containment leakage or loss of containment structural integrity. Other containment testing methods and inspections will assure all limiting conditions for operation will continue to be met. The margin of safety inherent in existing accident analyses will be maintained.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Attorney for licensee: Thomas C. Poindexter, Esq., Winston & Strawn, 1400 L Street, NW., Washington, DC 200053502.

NRC Section Chief: Stephen Dembek.
Entergy Nuclear Operations, Inc., Docket No. 50333, James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: June 22, 2004.

Description of amendment request: The proposed amendment would delete requirements from the Technical Specifications (TSs) to maintain hydrogen and oxygen monitors. A notice of availability for this technical specification improvement using the consolidated line item improvement process (CLIIP) was published in the Federal Register (FR) on September 25, 2003 (68 FR 55416). Licensees were generally required to implement upgrades as described in NUREG0737, ``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' and Regulatory Guide 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, ``Standards for combustible gas control system in light watercooled power reactors,'' eliminated the requirements for hydrogen recombiners (not installed at FitzPatrick 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 FR on September 25, 2003 (68 FR 55416). The licensee affirmed the applicability of the model NSHC determination in its application dated June 22, 2004.

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 [Regulatory Guide] 1.97 Category 1, is intended for key [[Page 53104]]
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 TS will not prevent an accident management strategy through the use of the severe accident management guidelines (SAMGs), the emergency plan (EP), the emergency operating procedures (EOPs), and site survey monitoring that support modification of emergency plan protective action recommendations (PARs).

Therefore, the relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, does not involve a significant increase in the probability or 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 relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, 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.

The relaxation of the hydrogen and oxygen monitor requirements, including removal of these requirements from TS, 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.

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 TS will not result in a significant reduction in their functionality, reliability, and availability.

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

Attorney for licensee: Mr. John Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Section Chief: Richard J. Laufer.
Entergy Nuclear Operations, Inc., Docket No. 50286, Indian Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: June 2, 2004.

Description of amendment request: The proposed amendment would revise the Technical Specifications (TSs) to fully adopt the alternate source term (AST) methodology for designbasis accident dose consequence evaluations in accordance with 10 CFR 50.67. Specifically, the amendment would revise the TS Definition regarding dose equivalent iodine and TS Section 5.5.10, ``Ventilation Filter Testing Program (VFTP).'' The AST methodology for the fuelhandling accident was previously approved in Amendment No. 215, dated March 17, 2003.

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

Response: No.

The proposed change involves the reanalysis of design basis radiological accidents in Containment and the Fuel Storage Building. The new analyses, based on the Alternate Source Term (AST), in accordance with 10 CFR 50.67, will replace the existing analyses that are based on the methodologies of [Atomic Energy Commission Report, ``Calculation of Distance Factors for Power and Test Reactor Sites,'' 1962] TID14844. As a result of the new analyses, changes to the Technical Specifications are proposed which take credit for the new analysis results.

The proposed changes to the Technical Specifications modify requirements regarding filter testing for a variety of systems (i.e., Containment Purge, Fuel Storage Building Emergency
Ventilation). The analyses do not credit charcoal or HEPA [high efficiency particulate air] filtration for dose mitigation. The proposed changes reflect the plant configuration that will support implementation of the AST analyses.

The AST analysis follows the guidance of the NRC Regulatory Guide 1.183 and uses the acceptance criteria of the NRC Standard Review Plan (NUREG0800) for offsite doses and General Design Criteria for Control Room personnel. The accident analyses conservatively assume that the Containment Building and the Fuel Storage Building, including ventilation filtration systems for those buildings, do not diminish or delay the assumed fission product release.

The proposed changes also revise the definition of Dose Equivalent Iodine (DEI) to be consistent with the assumptions of the analyses. The limits for DEI do not change as a result of the implementation of the AST analyses.

The change from the original source term to the new proposed AST is a change in analysis method and assumptions and has no effect on accident initiators or causal factors that contribute to the probability of occurrence of previously analyzed accidents. Use of AST to analyze the dose effect of design basis accidents shows that regulatory acceptance criteria for the new methodology continue to be met. Changing the analysis methodology does not change the sequence or progression of the accident scenario.

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

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

Response: No.

The changes proposed in this license amendment request involve the use of a new analysis methodology and related regulatory acceptance criteria. In addition, certain changes to plant ventilation systems can be made based on the analysis results, using the new methodology. Use of a new analysis
[[Page 53105]]
method does not impact the design or operation of plant systems or components and new accident scenarios would therefore not be created. The proposed changes to air ventilation and filtration systems do not adversely affect plant equipment used to protect plant safety limits or the way in which that plant equipment is operated or maintained. As a result, no new failure modes are being introduced that could lead to different accidents.

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 existing dose analysis methodology and assumptions demonstrate that the dose consequences for all design basis accidents are within regulatory limits for whole body and thyroid doses as established in 10 CFR 100 (except for the Fuel Handling Analysis, which is already based on the AST methodology). The alternate dose analysis methodology and assumptions also demonstrate that the dose consequences of these accidents are within the regulatory requirements established for the new methodology.

The limits applicable to the alternate analysis are established in 10 CFR 50.67 in conjunction with the Total Effective Dose Equivalent (TEDE) acceptance directed in Regulatory Guide 1.183. The acceptance criteria for both dose analysis methods have been developed for the purpose of evaluating design basis accidents to demonstrate adequate protection of public health and safety. An acceptable margin of safety is inherent in both types of acceptance criteria.

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

Attorney for licensee: Mr. John Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 10601.

NRC Section Chief: Richard J. Laufer.
Entergy Nuclear Operations, Inc., Docket No. 50286, Indian Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: June 3, 2004.

Description of amendment request: The proposed amendment would increase the maximum authorized reactor core power level from 3067.4 megawatt thermal (MWt) to 3216 MWt. This represents a nominal increase of 4.85% rated thermal power. The amendment would also revise the Technical Specifications (TSs) to relocate certain cyclespecific parameters to the Core Operating Limits Report (COLR) by adopting TS Task Force Traveler TSTF339, ``Relocate Technical Specification Parameters to the COLR.'' In addition, the amendment would revise several allowable values in TS Table 3.3.11, ``Reactor Protection System (RPS) Instrumentation,'' and Table 3.3.21, ``Engineered Safety Feature Actuation System (ESFAS) Instrumentation.''

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

Response: No.

The evaluations and analyses associated with this proposed change to core power level have demonstrated that all applicable acceptance criteria for plant systems, components, and analyses (including the Final Safety Analysis Report Chapter 14 safety analyses) will continue to be met for the proposed increase in licensed core thermal power for Indian Point 3 (IP3). The subject increase in core thermal power will not result in conditions that could adversely affect the integrity (material, design, and construction standards) or the operational performance of any potentially affected system, component or analysis. Therefore, the probability of an accident previously evaluated is not affected by this change. The subject increase in core thermal power will not adversely affect the ability of any safetyrelated system to meet its intended safety function. Further, the radiological dose evaluations in support of this power uprate effort show all acceptance criteria are met.

The relocation of cyclespecific core operating limits from the Technical Specifications to the Core Operating Limits Report (COLR), in accordance with TSTF339, has no influence or impact on the probability or consequences of a Design Basis Accident. Adherence to the COLR and accepted methodologies for establishing COLR parameters continues to be controlled by the plant Technical Specifications. Relocation of cyclespecific values to the COLR while maintaining the limiting requirements in the Technical Specifications reduces administrative burden associated with processing license amendments for routine core reload designs.

RPS and ESF [engineered safety feature] allowable values established in plant technical specifications represent acceptance criteria used by plant personnel in assessing the operability of instrumentation channels.

Allowable values are not accident initiators and have no role in the probability of occurrence of an accident. Safety analyses for design basis accidents use certain assumptions (Safety Analysis Limits) regarding the actuation of RPS and ESF protective functions. The proposed allowable values are developed using a methodology that assures the accident analysis assumptions are valid and the consequences of previously analyzed accidents continue to meet established limits.

Therefore, the proposed changes described in this license amendment request do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The analyses and evaluations performed for the proposed increase in power show that all applicable acceptance criteria for plant systems, components, and analyses (including FSAR [Final Safety Analysis Report] Chapter 14 safety analyses) will continue to be met for the proposed power increase in IP3 licensed core thermal power. The subject increase in core thermal power will not result in conditions that could adversely affect the integrity (material, design, and construction standards) or operational performance of any potentially affected system, component, or analyses. The subject increase in core thermal power will not adversely affect the ability of any safetyrelated system to meet its safety function.
Furthermore, the conditions and changes associated with the subject increase in core thermal power will neither cause initiation of any accident, nor create any new credible limiting single failure. The power uprate does not result in changing the status of events previously deemed to be noncredible being made credible.
Additionally, no new operating modes are proposed for the plant as a result of this requested change.

The relocation of cyclespecific core operating limits from the Technical Specifications to the Core Operating Limits Report (COLR), in accordance with TSTF339, does not involve any changes to plant equipment or the way is which the plant is operated. There are no new accident initiators or causal mechanisms being introduced by this proposed change. Relocation of cyclespecific values to the COLR while maintaining the limiting requirements in the Technical Specifications reduces administrative burden associated with processing license amendments for routine core reload designs.

RPS and ESF

SUMMARY:

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

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