ML070170268

Issuance of License Amendment 259 Maximum Allowable Power with Inoperable Main Steam Safety Valves

ML070170268
Person / Time
Site:	Salem
Issue date:	03/07/2007
From:	Richard Ennis NRC/NRR/ADRO/DORL/LPLI-2
To:	Levis W Public Service Enterprise Group
Ennis R, NRR/DORL, 415-1420
Shared Package
ML070170301	List: ML070170268 ML070680300
References
TAC MD1190
Download: ML070170268 (13)

Text

March 7, 2007 Mr. William Levis Senior Vice President & Chief Nuclear Officer PSEG Nuclear LLC - N09 Post Office Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 - ISSUANCE OF AMENDMENT RE: MAXIMUM ALLOWABLE POWER WITH INOPERABLE MAIN STEAM SAFETY VALVES (TAC NO. MD1190)

Dear Mr. Levis:

The Commission has issued the enclosed Amendment No. 259 to Facility Operating License No. DPR-75 for the Salem Nuclear Generating Station, Unit No. 2. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated April 6, 2006. The amendment changes the TSs to reduce the maximum allowable reactor power level when two main steam safety valves are inoperable.

A copy of our Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/ra/

Richard B. Ennis, Senior Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-311

Enclosures:

1. Amendment No. 259 to License No. DPR-75

2. Safety Evaluation cc w/encls: See next page

March 7, 2007 Mr. William Levis Senior Vice President & Chief Nuclear Officer PSEG Nuclear LLC - N09 Post Office Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 - ISSUANCE OF AMENDMENT RE: MAXIMUM ALLOWABLE POWER WITH INOPERABLE MAIN STEAM SAFETY VALVES (TAC NO. MD1190)

Dear Mr. Levis:

The Commission has issued the enclosed Amendment No. 259 to Facility Operating License No. DPR-75 for the Salem Nuclear Generating Station, Unit No. 2. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated April 6, 2006. The amendment changes the TSs to reduce the maximum allowable reactor power level when two main steam safety valves are inoperable.

A copy of our Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/ra/

Richard B. Ennis, Senior Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-311

Enclosures:

1. Amendment No. 259 to License No. DPR-75

2. Safety Evaluation cc w/encls: See next page DISTRIBUTION:

PUBLIC RidsOgcRp MVaaler LPL1-2 R/F RidsAcrsAcnwMailCenter RidsNrrDorlLpl1-2 RidsNrrDirsltsb RidsNrrLACRaynor RidsRgn1MailCenter RidsNrrPMREnnis GHill (2), OIS RidsNrrDorlDPR JNakoski Package Accession No.: ML070170301 Amendment Accession No: ML070170268 TS Accession No.: ML070680300

• By memo dated 9/13/06 OFFICE LPL1-2/PM LPL1-2/LA SPWB/BC ITSB/BC OGC LPL1-2/BC (wcomment)

NAME REnnis CRaynor JNakoski* TKobetz SUttal HChernoff DATE 1/25/07 1/19/07 9/13/06 1/23/07 3/2/07 3/7/07 OFFICIAL RECORD COPY

Salem Nuclear Generating Station, Unit Nos. 1 and 2 cc:

Mr. Dennis Winchester Township Clerk Vice President - Nuclear Assessment Lower Alloways Creek Township PSEG Nuclear Municipal Building, P.O. Box 157 P.O. Box 236 Hancocks Bridge, NJ 08038 Hancocks Bridge, NJ 08038 Mr. Paul Bauldauf, P.E., Asst. Director Mr. Thomas P. Joyce Radiation Protection Programs Site Vice President - Salem NJ Department of Environmental PSEG Nuclear Protection and Energy P.O. Box 236 CN 415 Hancocks Bridge, NJ 08038 Trenton, NJ 08625-0415 Mr. George H. Gellrich Mr. Brian Beam Plant Support Manager Board of Public Utilities PSEG Nuclear 2 Gateway Center, Tenth Floor P.O. Box 236 Newark, NJ 07102 Hancocks Bridge, NJ 08038 Regional Administrator, Region I Mr. Carl J. Fricker U.S. Nuclear Regulatory Commission Plant Manager - Salem 475 Allendale Road PSEG Nuclear - N21 King of Prussia, PA 19406 P.O. Box 236 Hancocks Bridge, NJ 08038 Senior Resident Inspector Salem Nuclear Generating Station Mr. Darin Benyak U.S. Nuclear Regulatory Commission Director - Regulatory Assurance Drawer 0509 PSEG Nuclear - N21 Hancocks Bridge, NJ 08038 P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. James Mallon Manager - Licensing 200 Exelon Way, KSA 3-E Kennett Square, PA 19348 Mr. Steven Mannon Manager - Regulatory Assurance P.O. Box 236 Hancocks Bridge, NJ 08038 Jeffrie J. Keenan, Esquire PSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038

PSEG NUCLEAR LLC EXELON GENERATION COMPANY, LLC DOCKET NO. 50-311 SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.259 License No. DPR-75

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment filed by PSEG Nuclear LLC, acting on behalf of itself and Exelon Generation Company, LLC (the licensees) dated April 6, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-75 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No. , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. The license amendment is effective as of its date of issuance and shall be implemented prior to restart from the steam generator replacement outage.

FOR THE NUCLEAR REGULATORY COMMISSION

/ra/

Harold K. Chernoff, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and the Technical Specifications Date of Issuance: March 7, 2007

ATTACHMENT TO LICENSE AMENDMENT NO. 259 FACILITY OPERATING LICENSE NO. DPR-75 DOCKET NO. 50-311 Replace the following page of Facility Operating License No. DPR-75 with the attached revised page as indicated. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert Page 4 Page 4 Replace the following page of the Appendix A, Technical Specifications, with the attached revised page as indicated. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert 3/4 7-2 3/4 7-2

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 259 TO FACILITY OPERATING LICENSE NO. DPR-75 PSEG NUCLEAR LLC EXELON GENERATION COMPANY, LLC SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 DOCKET NO. 50-311

1.0 INTRODUCTION

By letter dated April 6, 2006, PSEG Nuclear LLC (PSEG or the licensee) submitted a request for changes to the Salem Nuclear Generating Station (Salem), Unit No. 2, Technical Specifications (TSs). The proposed amendment would change the TSs to reduce the maximum allowable reactor power level when two main steam safety valves (MSSVs) are inoperable. Specifically, the amendment would revise TS Table 3.7-1, Maximum Allowable Power With Inoperable Steam Line Safety Valves, to reduce the allowable reactor power level from 59 percent to 58 percent of rated thermal power (RTP) when a maximum of two MSSVs are inoperable in any steam generator (SG). The proposed amendment supports the replacement of the SGs which is planned for the Salem Unit 2 sixteenth refueling outage (2R16) in the spring of 2008.

2.0 REGULATORY EVALUATION

The Nuclear Regulatory Commissions (NRCs or the Commissions) regulatory requirements related to the content of the TSs are set forth in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, Technical Specifications. This regulation requires that the TSs include items in five specific categories. These categories include (1) safety limits, limiting safety system settings and limiting control settings, (2) limiting conditions for operation, (3) surveillance requirements, (4) design features, and (5) administrative controls. Additionally, Criterion 2 of 10 CFR 50.36(c)(2)(ii) requires a limiting condition for operation be established for a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The licensees proposed change relates to Criterion 2 of 10 CFR 50.36(c)(2)(ii) in that Salem Unit 2 TS 3/4.7.1 includes restrictions on plant operation if MSSVs are inoperable in order to provide assurance that the secondary pressure limit of 110 percent of design pressure is met.

The proposed change to reduce the maximum RTP level with a maximum of two inoperable MSSVs in any SG establishes an operating restriction that is used as an initial condition of

transient analyses performed in support of the Salem Unit 2 SG replacement effort, which is planned to be completed during the spring 2008 refueling outage.

2.1 MSSVs Appendix A to 10 CFR Part 50, General Design Criterion (GDC) 15, Reactor Coolant System Design, states that the reactor coolant pressure boundary (RCPB) be designed, constructed, and tested with sufficient margin to ensure that design conditions are not exceeded during normal operation or anticipated operational occurrences. The overpressure protection system, of which the MSSVs are a part, is relied upon to maintain reactor coolant system (RCS) pressure within acceptable design limits during certain analyzed transients, and is required during Modes 1, 2, 3, 4, and 5. Overpressure protection for the RCPB is ensured by application of relief and safety valves as well as the reactor protection system.

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) requires that safety valves be designed with sufficient capacity to limit the pressure to less than 110 percent of the RCPB design pressure during the most severe abnormal operational transient with reactor scram. In addition, sufficient margin shall be available to account for uncertainties in the design and operation of the plant assuming:

(1) The reactor is operating at a power level that will produce the most severe overpressurization transient; (2) All system and core parameters are at values within normal operating range, including uncertainties and TS limits that produce the highest anticipated pressure; (3) The reactor scram is initiated by the second safety-grade signal from the reactor protection system; and (4) The discharge flow is based on the rated capacities specified in the ASME Code for each type of valve.

Full credit for actuation is allowed for spring-loaded safety valves designed in accordance with the requirements of the ASME Code. The chief purpose of the MSSVs is to provide overpressure protection for the secondary system. The MSSVs also provide protection against overpressurizing the RCPB by providing a heat sink for the removal of energy from the RCS if the preferred heat sink, provided by the condenser and circulating water system via the turbine bypass system, is not available.

At Salem Unit 2, five MSSVs are located on each main steam header outside containment and upstream of the main steam isolation valves. These valves are described Section 10.3.2.1 of the Salem Updated Final Safety Analysis Report (UFSAR). The MSSVs are set at 1070, 1100, 1110, 1120, and 1125 pounds per square inch gauge (psig) respectively, and have sufficient capacity, in conjunction with the power-operated relief valve, to limit the secondary system pressure to 110 percent of the SG design pressure (1085 psig) in order to meet the requirements of the ASME Code.

2.2 Inoperable MSSVs Operability of the MSSVs in accordance with Salem Unit 2 TS 3/4.7.1 ensures that the secondary system pressure will be limited to within 110 percent of its design pressure of 1085 psig during the most severe anticipated system operational transient. The events that challenge the relief capacity of the safety valves are those resulting in decreased heat removal capability. Analyses have shown that a loss of external electrical load and/or turbine trip (LOL/TT) is the limiting anticipated operational occurrence with respect to MSSV capability.

During the sixteenth refueling outage in the spring of 2008, the licensee plans to replace the four Salem Unit 2 original Westinghouse Series 51 SGs with AREVA/Framatome Model 61/19T SGs. As a result of analyses of the LOL/TT transient performed in support of the planned SG replacement, the values currently provided in TS 3/4.7.1, Table 3.7-1 for maximum allowable reactor power levels with inoperable MSSVs would no longer provide adequate overpressure protection or meet the current licensing basis acceptance criteria. Therefore, PSEG has requested changes to the allowable power levels provided in TS 3/4.7.1 in order to support the SG replacement effort. Specifically, for the case of two inoperable MSSVs, the RTP must be reduced by 1 percent (from 59 percent to 58 percent) to maintain peak secondary system pressure below the design limit of 1193.5 psig (110 percent of design pressure of 1085 psig).

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the proposed TS changes by reviewing: (1) the licensees application dated April 6, 2006; (2) applicable sections of the Salem UFSAR; (3) NUREG-1431, Revision 2, Standard Technical Specifications for Westinghouse Plants; (4) Technical Specifications Task Force (TSTF) TSTF-235, Revision 1; (5) NRC Information Notice (IN) 94-60, Potential Overpressurization of Main Steam System; and (6) NRC Letter to PSEG dated September 19, 2001, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Issuance of Amendment Re:

Reactor Coolant System Safety Valves and Plant Systems, Main Steam Safety Valves (TAC Nos. MB0087 and MB0088).

3.1 Determination of Limits Based on RTP Salem Unit 2 TS Action Statements 3.7.1.1.a and 3.7.1.1.b read as follows:

a. With one or two main steam line code safety valves inoperable in one or more steam generators, operation in Modes 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or reduce power to less than or equal to the applicable percent of RATED THERMAL POWER per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With three main steam line code safety valves inoperable in one or more steam generators, operation in Modes 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valves are restored to OPERABLE status or reduce power to less than or equal to the applicable percent of RATED THERMAL POWER per Table 3.7-1 and within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, reduce the Power Range Neutron Flux High trip setpoint to less than or equal to the RATED THERMAL POWER

per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />."

Currently, Salem Unit 2 TS Table 3.7-1 establishes the following allowable reactor power limits when MSSVs are inoperable:

Maximum Number of Maximum Allowable Power*

Inoperable MSSVs (Percent of RTP) on Any Operating SG 1 87 2 59 3 39

• These values do not provide any allowance for calorimetric error.

In the September 19, 2001, amendment to the Salem Unit 2 Facility Operating License (Amendment No. 225), PSEG was given permission to prescribe the current allowable power levels based on RTP when one or two MSSVs are inoperable without concurrently reducing the power range neutron flux scram setpoint. These new requirements were similar to, but not exactly the same as, the required actions provided in the Westinghouse improved standard TSs, NUREG-1431, Revision 2.

In general, the required limitations on RTP necessary to prevent secondary system overpressurization may be determined by system transient analysis or by a heat balance calculation. In certain circumstances it may be necessary to reduce the setpoint of the power range neutron flux - high reactor trip function in order to limit the generation of heat within the primary system during an abnormal operational occurrence. NUREG-1431, Revision 2, allows continued operation at a reduced power level with one inoperable MSSV without requiring the power range neutron flux high trip setpoint to be reduced. However, Salem Unit 2 was given permission to operate at a reduced power level with two inoperable MSSVs without a subsequent change in the power range neutron flux high trip setpoint. This was allowed because, in certain cases when the moderator temperature coefficient is zero or negative at all power levels, as is the case for Salem Unit 2, an increase in primary (moderator) temperature would add sufficient negative reactivity to cause reactor power to decrease to levels that would further allow the RCS and main steam system pressure to remain below their respective ASME Code allowable limits.

PSEGs analysis to substantiate this point was consistent with guidance provided in NRC IN 94-60. IN 94-60 alerted licensees to consider recommendations provided in Westinghouse Nuclear Safety Advisory Letter (NSAL)94-001, "Operation at Reduced Power Levels With Inoperable MSSVs," dated January 20, 1994. NSAL 94-001 recommended the use of more conservative methods to determine power range high neutron flux trip setpoints.

In the case of one or more SGs having three MSSVs declared inoperable, PSEG is required to reduce power to 39 percent within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and decrease the power range neutron flux high trip setpoint within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The completion time of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> is consistent with NUREG-1431, Revision 2, and TSTF-235, Revision 1, which provides a reasonable time to correct the MSSV

inoperability, and is based on: (1) time to perform the power reduction; (2) operating experience in resetting all channels of a protective function; and (3) the low probability of the occurrence of a transient that could result in SG overpressure during this period.

3.2 Allowable Power Level With Two Inoperable MSSVs As previously stated, an LOL/TT is the limiting anticipated operational occurrence with respect to MSSV capability. LOL/TT is the loss of steam load during power operation without a direct reactor trip. The licensee analyzed LOL/TT to verify that: (1) pressure in the RCS and main steam system would remain below 110 percent of design pressure; (2) minimum departure from nucleate boiling ratio remains above the analysis limit, thereby assuring fuel cladding integrity is maintained; and (3) the pressurizer does not fill (i.e., the LOL/TT event does not result in a more serious plant condition).

After a reanalysis conducted by Westinghouse of the LOL/TT transient in preparation for the replacement of the Salem Unit 2 Westinghouse Series 51 original steam generators with AREVA/Framatome Model 61/19T replacement steam generators (RSGs), the licensee concluded that a reduction in allowable reactor power is necessary to meet the secondary system pressure limit of 110 percent of design pressure during a LOL/TT event with a maximum of two MSSVs inoperable on any SG.

The need for the proposed reduction in RTP is attributed to differences in SG geometry and operating conditions for the RSGs. Specifically, primary side volume is greater for the RSGs, secondary heat transfer rate is greater for the RSGs, and main steam pressures are higher with the RSGs.

Westinghouse analyzed the LOL/TT transient in accordance with the Standard Thermal Design Procedure, which is used for analysis cases that demonstrate the adequacy of the pressure relieving devices. This procedure requires that all variables are assumed to be at their maximum values consistent with steady-state full power operation, resulting in the maximum power difference for the load loss. In addition, the analyses of the LOL/TT were performed using the NRC-approved LOFTRAN computer code, modeling the RSGs and using assumptions consistent with the current licensing basis for operation with inoperable MSSVs as previously approved for Salem Unit 2. These assumptions tend to maximize primary to secondary heat transfer during the LOL/TT event and, therefore, result in a conservative calculation of maximum secondary pressure.

The LOL/TT analyses show that with a maximum of two inoperable MSSVs in any SG, operation at 58 percent RTP would result in a maximum secondary pressure of 1190.7 psig, which is below the acceptance criterion of 1193.5 psig. Therefore, the proposed reduction in RTP with two inoperable MSSVs from 59 percent to 58 percent will be sufficient to maintain peak secondary system pressure below the limit Based on the proposed allowable values for Table 3.7-1 which were calculated using plant-specific analyses for Salem Unit 2 that: (1) are in accordance with previously-approved NRC methodologies; and (2) reduce the current values in the TSs to levels that would preclude SG overpressurization with the RSGs in place, the NRC staff concludes that the proposed values are acceptable.

3.3 Rod Withdrawal at Power Transient Analysis The rod withdrawal at power (RWAP) transient with inoperable MSSVs can likewise pose a challenge to the capacity of the remaining MSSVs. The licensee analyzed this event to verify that sufficient MSSV capacity is available to mitigate the reactivity transient during operation consistent with TS Table 3.7-1. Analysis cases were performed at each of the TS Table 3.7-1 power levels using a range of reactivity insertion rates from 1 percent millrho (pcm)/second to 110 pcm/second.

The RWAP analyses were performed under the same NRC-approved methods as the LOL/TT analyses discussed previously, and demonstrate that the acceptance criteria are met with the current TS limits. Therefore, operation in accordance with the proposed RTP change for two inoperable MSSVs is conservative with respect to the analysis.

3.4 Changes to the MSSV Capacity The licensee noted in its submittal that TS Bases Section 3/4.7.1 provides a comparison of maximum calculated steam flows to MSSV capacity. The total relieving capacity of 16.66 E6 pounds-mass per hour (Ibm/hr) is 110.4 percent of the calculated maximum flow of 15.08 E6 Ibm/hr (current TS Bases value). Calculated steam flow for the RSGs at the upper Tavg limit (577.9 degrees F), 100-percent reactor power (3459 megawatts thermal) and 0 percent SG tube plugging, is 15.12 E6 Ibm/hr. Steam flow capacity remains at approximately 110 percent of the maximum value and is not adversely impacted by the proposed change.

3.5 Technical Evaluation Conclusion The proposed amendment reduces the maximum allowable power level with two inoperable MSSVs from 59 percent to 58 percent in support of the Salem Unit 2 SG replacement effort that is planned for the spring 2008 refueling outage. This change will ensure that the secondary side pressure limit is not exceeded during the worst case transient event. Based on the proposed allowable values being calculated using plant-specific analyses for Salem Unit 2 that:

(1) are in accordance with previously approved NRC methodologies, and (2) reduce the current values in the TSs to levels that would preclude SG overpressurization with the RSGs in place, the NRC staff finds that the proposed changes are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New Jersey State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding

(71 FR 65144). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: M. Vaaler Date: March 7, 2007