ML081190469

Correction to Amendment No. 311 for Core Operating Limits Report References for Realistic Large Break Loss-of-Coolant Accident Methodology

ML081190469
Person / Time
Site:	Sequoyah
Issue date:	07/03/2008
From:	Moroney B NRC/NRR/ADRO/DORL/LPLII-1
To:	Campbell W Tennessee Valley Authority
Moroney B, NRR/DORL, 415-3974
References
TAC MD6259
Download: ML081190469 (13)

Text

July 3, 2008 Mr. William R. Campbell, Jr.

Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

Subject:

SEQUOYAH NUCLEAR PLANT, UNIT 2 - CORRECTION TO AMENDMENT NO. 311 FOR CORE OPERATING LIMITS REPORT REFERENCES FOR REALISTIC LARGE BREAK LOSS-OF-COOLANT ACCIDENT METHODOLOGY (TAC NO. MD6259)

Dear Mr. Campbell:

On April 10, 2008, the U.S. Nuclear Regulatory Commission (NRC) issued Amendment No. 311 to Facility Operating License No. DPR-79 for the Sequoyah Nuclear Plant, Unit 2 (Agencywide Documents Access and Management System Accession Number ML080810353).

Subsequently, the Sequoyah licensing staff and its contractor AREVA informed the NRC staff of inaccuracies in the supporting safety evaluation (SE). Specifically, rated thermal power was stated as 3479 megawatt thermal (Mwt) instead of 3455 Mwt, and there was a need to clarify that 3479 Mwt is the maximum power used in the accident analysis. Also, Appendix K to Title 10, Code of Federal Regulations, Part 50, was incorrectly listed as a regulatory requirement for the evaluation.

The SE has been corrected to reflect these comments and make associated additional administrative changes. The enclosed SE should replace the one previously issued. The corrections do not affect the conclusions of the original SE.

We appreciate the constructive feedback from the TVA staff and regret any inconvenience this may have caused. If you have any further concerns, please contact me at 301-415-3974.

Sincerely,

/ra/

Brendan T. Moroney, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-328

Enclosure:

Revised SE cc w/encl: See next page

William R. Campbell, Jr.

SEQUOYAH NUCLEAR PLANT Tennessee Valley Authority cc:

Mr. Ashok S. Bhatnagar Senior Vice President Nuclear Generation Development and Construction Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Vice President Nuclear Support Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Michael J. Lorek Vice President Nuclear Engineering & Projects Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Timothy P. Cleary, Site Vice President Sequoyah Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Soddy Daisy, TN 37384-2000 General Counsel Tennessee Valley Authority 6A West Tower 400 West Summit Hill Drive Knoxville, TN 37902 Mr. John C. Fornicola, Manager Nuclear Assurance Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Ms. Beth A. Wetzel, Manager Corporate Nuclear Licensing and Industry Affairs Tennessee Valley Authority 4K Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801 Mr. James D. Smith, Manager Licensing and Industry Affairs Sequoyah Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Soddy Daisy, TN 37384-2000 Mr. Christopher R. Church, Plant Manager Sequoyah Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Soddy Daisy, TN 37384-2000 Senior Resident Inspector Sequoyah Nuclear Plant U.S. Nuclear Regulatory Commission 2600 Igou Ferry Road Soddy Daisy, TN 37379 Mr. Lawrence E. Nanney, Director TN Dept. of Environment & Conservation Division of Radiological Health Third Floor, L and C Annex 401 Church Street Nashville, TN 37243-1532 County Mayor Hamilton County Courthouse Chattanooga, TN 37402-2801 Mr. Larry E. Nicholson, General Manager Performance Improvement Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Michael A. Purcell Senior Licensing Manager Nuclear Power Group Tennessee Valley Authority 4K Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801 Ms. Ann P. Harris 341 Swing Loop Road Rockwood, TN 37854

July 3, 2008 Mr. William R. Campbell, Jr.

Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

Subject:

SEQUOYAH NUCLEAR PLANT, UNIT 2 - CORRECTION TO AMENDMENT NO. 311 FOR CORE OPERATING LIMITS REPORT REFERENCES FOR REALISTIC LARGE BREAK LOSS-OF-COOLANT ACCIDENT METHODOLOGY (TAC NO. MD6259)

Dear Mr. Campbell:

On April 10, 2008, the U.S. Nuclear Regulatory Commission (NRC) issued Amendment No. 311 to Facility Operating License No. DPR-79 for the Sequoyah Nuclear Plant, Unit 2 (Agencywide Documents Access and Management System Accession Number ML080810353).

Subsequently, the Sequoyah licensing staff and its contractor AREVA informed the NRC staff of inaccuracies in the supporting safety evaluation (SE). Specifically, rated thermal power was stated as 3479 megawatt thermal (Mwt) instead of 3455 Mwt, and there was a need to clarify that 3479 Mwt is the maximum power used in the accident analysis. Also, Appendix K to Title 10, Code of Federal Regulations, Part 50, was incorrectly listed as a regulatory requirement for the evaluation.

The SE has been corrected to reflect these comments and make associated additional administrative changes. The enclosed SE should replace the one previously issued. The corrections do not affect the conclusions of the original SE.

We appreciate the constructive feedback from the TVA staff and regret any inconvenience this may have caused. If you have any further concerns, please contact me at 301-415-3974.

Sincerely,

/ra/

Brendan T. Moroney, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-328

Enclosure:

Revised SE cc w/encl: See next page DISTRIBUTION:

PUBLIC RidsNrrDorlLpl2-2 RidsOgcRp RidsNrrDorlDpr RidsAcrsAcnw&mMailCenter RidsNrrDirsItsb AObodoako RidsNrrPMBMoroney RidsRgn2MailCenter RidsNrrDssSrxb FOrr RidsNrrLACSola RidsNrrDssScvb LPL2-2 R/F RLobel ADAMS Accession Number: ML081190469 NRR-058 OFFICE LPL2-2/PM LPL2-2/LA SRXB/BC SCVB/BC LPL2-2/BC NAME BMoroney RSola GCranston RLobel for RDennnig TBoyce DATE 7/2/08 7/2/08 7/2/08 5/29/08 7/3/08 OFFICIAL RECORD COPY

Correction Letter dated July 3, 2008 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 311 TO FACILITY OPERATING LICENSE NO. DRP-79 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 2 DOCKET NO. 50-328

1.0 INTRODUCTION

By letter dated July 26, 2007 (Ref. 1), the Tennessee Valley Authority (TVA), the licensee for Sequoyah Nuclear Plant, Unit 2 (SQN-2) requested an amendment to the SQN-2 operating license to add a new reference in Technical Specification (TS) Section 6.9.1.14.a. The new reference is AREVA Topical Report EMF-2103P-A, Realistic Large Break LOCA

[Loss-Of-Coolant Accident] Methodology for Pressurized Water Reactors (EMF-2103P-A). to Ref. 1 provided the SQN-2 evaluation of the large break LOCA (LBLOCA), as documented in AREVA Topical Report ANP-2655P, Revision 0, dated June 2007 (ANP-2655P).

The licensee supplemented this request by letters dated October 3, 2007 (Ref. 2), and December 21, 2007 (Ref. 3), and February 29, 2008 (Ref. 4). Ref. 4 provided Revision 1 to ANP-2655P, dated February 2008. The supplemental letters provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register.

The U.S. Nuclear Regulatory Commission (NRC or Commission) staff reviewed the licensees demonstration evaluations of the emergency core cooling system (ECCS) performance analyses, done in accordance with the AREVA best estimate (BE) LBLOCA methodology, with SQN-2 operating at its currently licensed core power of 3455 megawatt thermal (Mwt). These specific analyses were performed to demonstrate that the SQN-2 plant meets NRC requirements and the criteria of Section 50.46 of Title 10 of the Code of Federal Regulations (10 CFR 50.46). Also, these specific analyses when approved herein will be acceptable and, specifically, applicable to SQN-2 operated with the fuels identified in Table 1 of this SE. The BE LBLOCA analyses for SQN-2 were conducted assuming that the plant uses cores containing M5 clad uranium oxide fuel assemblies.

The NRC staff also reviewed the calculation of a minimum containment pressure during the postulated LOCA.

Correction Letter dated July 3, 2008

2.0 REGULATORY EVALUATION

2.1 Containment Analysis Standard Review Plan, Section 6.2.1.5, Minimum Containment Pressure Analysis for Emergency Core Cooling System Performance Capability Studies, provides guidance on assumptions made regarding the operation of engineered safety feature containment heat removal systems and the effectiveness of structural heat sinks within containment to remove energy from the containment atmosphere. The objective of this guidance is to obtain a conservatively large heat removal rate from the containment atmosphere in order to minimize the containment pressure. The accompanying Branch Technical Position CSB 6-2, Minimum Containment Pressure Model for PWR [Pressurized Water Reactor] ECCS Performance Evaluation, provides detailed guidance to achieve the same objective. The licensee has treated some parameters using the guidance of CSB 6-2 and other parameters statistically.

Regulatory Guide 1.157 (issued May 1989), Best Estimate Calculations of Emergency Core Cooling System Performance, which the licensee uses as guidance for the RLBLOCA analyses, states that:

The containment pressure used for evaluating the post-blowdown phase of the loss-of-coolant accident should be calculated in a best estimate manner and should include the effects of containment heat sinks. The calculation should include the effects of operation of all pressure reducing equipment assumed to be available. Best estimate models will be considered acceptable provided their technical basis is demonstrated with appropriate data and analysis. (Section C, Paragraph 3.12.1)

Regulatory Guide 1.157 also states (Section C.1) that the introduction of conservative bias is acceptable as long as this does not result in unrealistic results or omit important phenomena.

2.2 LBLOCA Analysis The BE LBLOCA analyses were performed to demonstrate that the ECCS design would provide sufficient ECCS flow to transfer the heat from the reactor core following a LBLOCA at a rate such that: (1) fuel and clad damage that could interfere with continued effective core cooling would be prevented, and (2) the clad metal-water reaction would be limited to less than the amounts that would compromise cladding ductility and result in excessive hydrogen generation.

The NRC staff reviewed the analyses to assure that the safety functions could be accomplished, assuming a single failure, for LBLOCAs and considering the availability of only onsite or offsite electric power (i.e., assuming offsite electric power is not available, with onsite electric power available; or assuming onsite electric power is not available, with offsite electric power available).

The NRC staff used the acceptance criteria for ECCS performance provided in 10 CFR 50.46 in assessing the application of the AREVA RLBLOCA methodology for SQN-2. In its assessment of the application of the methodology for SQN-2, the NRC staff also reviewed the limitations and Correction Letter dated July 3, 2008 conditions stated in its SE (Ref. 5) supporting general approval of the AREVA RLBLOCA methodology and the range of parameters described in the AREVA RLBLOCA methodology topical report.

3.0 TECHNICAL EVALUATION

3.1 Evaluation of Containment Pressure Calculation The SQN-2 RLBLOCA analysis assumes a break in the cold leg piping between the reactor coolant pump and the reactor vessel for the reactor coolant system loop containing the pressurizer. ANP-2655P states that this assumption is based on deterministic studies.

Considering the assumed single failure, one charging pump, one safety injection pump and one residual heat removal pump are assumed to be operating. In addition, ANP-2655P states that:

Regardless of the single-failure assumptions, all containment pressure reducing systems are assumed functional.

This complies with the guidance of NRC Branch Technical Position 6-2 and is acceptable since this reduces the containment accident pressure.

The SQN-2 Updated Final Safety Analysis Report (UFSAR), Section 6.2.2.1, states that the SQN-2 containment heat removal system consists of the following:

1. Ice Condenser

2. Air Return Fan System

3. Containment Spray System

4. Residual Heat Removal Spray System Of these, only the ice condenser and the containment spray system are important in the time frame of the calculation of the peak cladding temperature (PCT) and core quenching. (The SQN-2 design does not include fan coolers in its building pressure control.)

The assumed initial ice mass is 2.448 million pounds. The licensee characterizes this as a nominal ice mass. (It is 10 percent more than the minimum amount allowed by the SQN-2 TSs.)

The licensee states that:

Less than half of the ice (approximately 1.2 million pounds) is calculated to melt during the LOCA transients (i.e., the first 500 seconds of the transient).

Therefore, the ice mass used in the ICECON simulation is adequate for the prediction of containment pressures (break back pressures) during [the] LOCA.

The staff agrees with this assumption.

The start time of the air recirculation fan system is 600 seconds. The licensee states, in response to a staff request for additional information, that in all of the RLBLOCA cases, core Correction Letter dated July 3, 2008 quench occurred in less than 500 seconds. Therefore, the air return fan system is not significant in these analyses.

As stated above, both trains of the containment spray system are assumed to operate. The containment spray pumps take suction from the refueling water storage tank (RWST). The RWST water temperature is conservatively assumed to be 55 degrees Fahrenheit (°F).

(The RWST water temperature is taken as 110 °F for ECCS injection into the vessel.) In addition, the licensee assumes an early start to the containment spray flow (eight seconds after the containment safety injection signal). Both pumps are activated at 100 percent capacity (7700 gallons per minute).

The residual heat removal spray system is only needed after ice melt has occurred and is therefore not a significant factor in these analyses. Since the containment spray pump model is conservative (both trains operating, low water temperature, early start of flow), the NRC staff finds the containment spray modeling to be acceptable.

The NRC has previously approved the realistic LOCA containment modeling for a PWR with a subatmospheric containment (Ref. 6). In response to a staff request for additional information, the licensee stated that the general aspects of the approach approved by the NRC for the subatmospheric containment are the same as those for SQN-2. This approach uses both realistic and conservative modeling assumptions. The dominant phenomena influencing the containment response are the initial pressure and volume, heat transfer to internal structures, break size and effluent modeling. The break size and effluent modeling are not addressed in this SE report input. The phenomena identification and ranking table analyses performed by AREVA indicate that two containment-related phenomena directly influence PCT: the initial containment pressure and initial containment temperature (Ref. 1).

The containment pressure is calculated using values of the containment volume which are sampled over a range from a nominal volume to the containment empty volume. The containment empty volume is the volume with no internal equipment or structures included.

Since the pressure decreases with increasing containment volume, this range ensures that the selection of the containment volume is either best estimate (if the nominal containment volume is sampled) or conservative (that is, results in a pressure lower than nominal). The initial containment atmosphere temperature ranges between 80 °F to 100 °F for the upper compartment, and 95 °F to 130 °F for the lower compartment. The temperature is important in determining the air mass within the containment which affects the containment pressure. These temperature ranges are typical and acceptable. Both variables are assumed to have uniform distributions over their respective ranges.

The licensee derives the necessary information for the passive heat sinks from the SQN-2 minimum containment pressure model approved by the NRC for use with the deterministic LOCA evaluation performed in accordance with 10 CFR 50, Appendix K. The licensee states that the surface area of these heat structures was increased by five percent to provide further margin toward low pressures. Since these data are conservative for containment minimum pressure calculations, they are acceptable for the realistic LOCA calculations since the guidance of Regulatory Guide 1.157 permits conservative input as long as the use of such Correction Letter dated July 3, 2008 information does not make the analysis unrealistic. It is the NRC staffs judgment that this is the case.

The heat transfer coefficients for determining the heat transfer between the containment atmosphere and the passive heat sinks are important. The realistic LOCA methods use the Uchida condensing heat transfer correlation with a multiplier to account for the greater amount of heat transfer which would occur during the blowdown phase of the LOCA. The NRC staff has previously found this acceptable when justified on a plant specific basis since the value of the multiplier could change with each analysis. The licensee stated that further conservatism was provided as follows (Ref. 3):

Surface coatings, where they existed, were modeled as the underlying material so that the insulating effect on a metallic structure was discounted. As a further pressure lowering measure, condensing heat transfer was modeled on those structures which fall below the transient water level during the LOCA.

An important part of the containment pressure calculation for an ice condenser containment such as SQN-2s is the heat transfer between the containment steam and the ice melt. The licensee states that the water spillage plus drainage from the ice chest falls through the lower containment vapor. This condenses steam and reduces the containment pressure. The ice chest drainage flow is treated as 100 percent efficient spray during the post-blowdown period of the transient.

This modeling is conservative and acceptable.

The containment pressure is calculated by S-RELAP using containment models derived from the ICECON computer code. The NRC staff has previously approved these methods for minimum pressure calculations (Ref. 6). It is noted that ICECON is based on the CONTEMPT computer code. An NRC staff study dated August 31, 2002, has shown that the CONTEMPT code containment spray model tends to reduce pressure more rapidly than the data indicates

[sic]. Therefore, the containment spray model is acceptable for containment pressure calculations.

The staff finds the licensees referencing and use of the methods of EMF-2103P-A acceptable with respect to the determination of containment pressure since the calculation models have been previously approved, the calculations use best estimate or conservative input values and models, the licensee has demonstrated the acceptability of the multiplier used with the Uchida heat transfer correlation and, in general, the containment pressure calculation is in conformance with the guidance of Regulatory Guide 1.157 and the calculations are not so conservative as to be unrealistic.

3.2 Evaluation of AREVA BE LBLOCA Analyses Methodology In Ref. 3, the licensee stated, TVA and the LBLOCA Analysis Vendor have ongoing processes to ensure that all input variables and parameter ranges analyses for the SQN-2 RLBLOCA are verified as conservative with respect to plant operating design conditions revisions and updates to the analysis parameters are documented and approved in accordance with the processes Correction Letter dated July 3, 2008 described above for the initial analysis. Also, both Rev. 0 and Rev. 1 to ANP-2655 (page 3-6) contain the statement, Both entities have ongoing processes that assure the ranges and values of input parameters for the Sequoyah Unit 2 Station RLBLOCA analysis bound those of the as-operated plant. The staff finds that these statements, along with the generic acceptance of the AREVA RLBLOCA analysis methodology, provide assurance that the AREVA RLBLOCA analysis methodology and its LBLOCA analyses apply to SQN-2 operated at its current licensed power level.

In Ref. 1, the licensee provided the results for the SQN-2 RLBLOCA analyses, operating at the rated power of 3455 Mwt (performed in accordance with the AREVA RLBLOCA methodology).

These results were updated in Ref. 4. The licensees results for the calculated PCTs, the maximum cladding oxidations (local), and the maximum core-wide cladding oxidations for SQN-2 are provided in the following table along with the acceptance criteria of 10 CFR 50.46(b).

TABLE 1 SQN-2 LBLOCA ANALYSIS RESULTS (From Ref. 4, Table 3-6)

Parameter SQN-2 RLBLOCA Results 10 CFR 50.46 Limits Limiting Break Size/Location 2.7259 ft2/side Split/PD N/A Cladding Material M-5 (Cylindrical) Zircaloy, Zirlo, (or M-5**)

Peak Clad Temperature 2002 °F 2200 °F (10 CFR 50.46(b)(1))

Maximum Local Oxidation 3.42 percent 17.0 percent (10 CFR 50.46(b)(2))

Maximum Total Core-Wide Oxidation (All Fuel)

< 0.02 percent 1.0 percent (10 CFR 50.46(b)(3))

• Split/PD is a split break at the pump discharge.

• • M-5 is equivalent to Zircaloy and ZIRLO in applications of 10 CFR 50.46 criteria.

The licensees analysis takes credit for the increased accuracy of a leading edge flow meter device in the determination of the reactor operating power assumed in the LBLOCA analyses.

The power assumed in the analyses (3479 Mwt) is 0.7 percent higher than the rated power (3455 Mwt) to account for measurement inaccuracies.

The licensees LBLOCA analyses for SQN-2 properly assumed SQN-2 operating at a constant core power, and did not range reactor core power. The NRC staff finds this acceptable.

The analyses addressed the availability of offsite power correctly by ranging each case separately. This is acceptable because it satisfies General Design Criterion 35 of 10 CFR 50 Appendix A, in that each distribution type has been accounted for separately with its own set of Correction Letter dated July 3, 2008 cases, thereby addressing possible concerns associated with the mixing of two separate statistical spectra. Therefore, the NRC staff finds this treatment acceptable as discussed.

On August 3, 1998, the NRC issued Information Notice 98-29: Predicted Increase in Fuel Rod Cladding Oxidation, expressing concern regarding predicted cladding total oxidation (including both pre-accident and accident oxidation) resulting from a postulated LOCA could for some plants exceed the 17 percent limit, and that LOCA methodologies were not addressing that concern.

In letters dated March 31 and November 8, 1999, to Nuclear Energy Institute, NRC provided its position that both pre-accident and accident oxidation must be estimated, citing several references, including the Opinion of the Commission dated December 28, 1973, that demonstrate that the NRC position regarding oxidation predates the present LOCA acceptance criteria and the first accepted LOCA evaluation models under those criteria.

Therefore, the NRC staff considers the licensees estimate of oxidation resulting from the postulated LBLOCA alone low for M-5 cladding (3.42 percent, without considering pre-LOCA oxidation and without considering oxidation on both inside and outside surfaces). Also, the low calculated oxidation level is, at least in part, attributable to the slightly lower than anticipated calculated LBLOCA PCT (2002 °F). The low calculated total core-wide oxidation for LBLOCA is also most likely the result of the LBLOCA analysis methodologys consideration of only fresh fuel assemblies in the analyses and the treatment of radiation heat transfer.

However, the NRC staff also considers that, even if the calculated PCT and local oxidation were reasonably greater (less than 20 °F and 10 percent greater, respectively, see below) than reported above, the total expected oxidation, including all factors which should have been considered, would most likely be less than the 17 percent limit specified in the regulation.

The concern with core-wide oxidation relates to the amount of hydrogen generated during a LOCA. Because hydrogen that may have been generated pre-LOCA (during normal operation) will be removed from the reactor coolant system throughout the operating cycle, the NRC staff noted that pre-existing oxidation does not contribute to the amount of hydrogen generated post-LOCA and, therefore, it does not need to be addressed further when determining whether the calculated total core-wide oxidation meets the 1.0 percent criterion of 10 CFR 50.46(b)(3).

As discussed previously, TVA had AREVA conduct BE LBLOCA analyses for SQN-2 operating at the current licensed power level of 3455 Mwt, using an NRC-approved AREVA BE LBLOCA analysis methodology. In consideration of SQN-2 analyses regarding post-LOCA downcomer boiling, the NRC staff concluded that the PCT results of these analyses indicate compliance with 10 CFR 50.46(b)(1) through (b)(3) for licensed power levels of up to 3455 Mwt. Meeting these criteria provides reasonable assurance that at the current licensed power level, the SQN-2 core will be amenable to cooling as required by 10 CFR 50.46(b)(4). The capability of SQN-2 to satisfy the long term cooling requirements of 10 CFR 50.46(b)(5) is unaffected by this amendment.

Correction Letter dated July 3, 2008 3.3 SQN-2 TS Change The licensee proposed to revise TS 6.9.1.14.a to reflect use of a new LBLOCA analysis methodology to perform LBLOCA analyses in support of SQN-2 operation. Specifically, TS Page 6-14, CORE OPERATING LIMITS REPORT (continued), would be amended to add EMF-2103P-A as the licensing basis LBLOCA methodology for the SQN-2 plant. The NRC staff reviewed the TS provision, assessed it for consistency against NUREG-1431, Revision 3 (Ref. 7), as stated below, and found its content acceptable and compatible with a proposed COLR.

For the proposed TSs change, the following reference will be added:

9.

EMF-2103P-A, Realistic Large Break LOCA Methodology for Pressurized Water Reactors.

This methodology was found to apply to all conventional Westinghouse and Combustion Engineering PWR designs in the NRC generic SE of the EMF-2103P-A, Rev. 0 methodology.

Therefore, the realistic LBLOCA methodology described in EMF-2103P-A, Rev. 0 is acceptable for application to SQN-2, which is a PWR of Westinghouse design, and for inclusion in TS for the SQN-2 plant. The proposed change does not include the EMF-2103P-A revision number; nor does it include the date of approval for the methodology. The licensee will list the topical report, including the latest revision number (used at SQN-2) and date of NRC approval, in the COLR for SQN-2, consistent with guidance provided in NUREG-1431.

The NRC staff finds that the EMF-2103P-A methodology (with the modifications made to address staff review issues) is applicable to SQN-2, and that the limitations and conditions of the NRCs SE approving the EMF-2103P-A methodology were satisfied, for the present SQN-2 operating power. The staff concludes that the proposed addition of EMF-2103P-A to SQN-2 TS 6.9.1.14.a is acceptable.

In summary, based on the above discussion, the NRC staff concluded that the AREVA BE LBLOCA analyses methodology, as described in EMF-2103P-A, Rev. 0 is acceptable for use by SQN-2 in demonstrating compliance with the requirements of 10 CFR 50.46(b). The NRC staffs conclusion is based on the staffs verification that the SQN-2 plants design is among the designs for which AREVA BE LBLOCA application was approved.

The NRC staffs review of the acceptability of the AREVA BE LBLOCA methodology for SQN-2 focused on assuring that the licensee and its vendor have processes to assure that specific input parameters or bounding values and ranges (where appropriate) are used to conduct the SQN-2 LBLOCA analyses, that the analyses will be conducted within the conditions and limitations of the NRC-approved AREVA BE LBLOCA methodology, and that the results will satisfy the requirements of 10 CFR 50.46(b) for SQN-2 operating at its present licensed power.

This SE also documents the NRC staff review and acceptance of the AREVA BE LBLOCA BE LBLOCA analysis methodology for application to SQN-2, for inclusion in the SQN-2 TSs and COLR, and of the specific LBLOCA analyses discussed above that were performed with the Correction Letter dated July 3, 2008 AREVA BE LBLOCA methodology for SQN-2 operated at powers up to its licensed power level of 3455 Mwt, and that the use of the AREVA fuel would not adversely affect calculated ECCS performance in response to LOCA events.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Tennessee State official, Mr. Bruce House of the Tennessee Bureau of Radiological Health, 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 (72 FR 49583). 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.

7.0 References

1.

Letter from Glenn W. Morris, Manager, Site Licensing and Industry Affairs, TVA, to NRC, Sequoyah Nuclear Plant, Unit 2 - Technical Specifications (TS) Change 07-04, Revision of Core Operating Limits Report (COLR) References for Realistic Large Break Loss of Coolant Accident Methodology, July 26, 2007.

2.

Letter from Glenn W. Morris, Manager, Site Licensing and Industry Affairs, TVA, to NRC, Sequoyah Nuclear Plant, Unit 2 - Technical Specifications (TS) Change 07-04, Revision of Core Operating Limits Report (COLR) References for Realistic Large Break Loss of Coolant Accident Methodology Supplemental Information, October 3, 1997.

3.

Letter from James D. Smith, Manager, Site, Licensing and Industry Affairs, TVA, to NRC, Sequoyah Nuclear Plant, Unit 2 - Response to Request for Additional information (RAI)

Correction Letter dated July 3, 2008 Regarding Large Break Loss-of-Coolant Accident Analysis Methods, December 21, 2007.

4.

Letter from James D. Smith, Manager, Site, Licensing and Industry Affairs, TVA, to NRC, Sequoyah Nuclear Plant, Unit 2 - Response to Request for Additional information (RAI)

Regarding Large Break Loss-of-Coolant Accident Analysis Methods, February 29, 2008.

5.

NRC Safety Evaluation, Safety Evaluation on Framatome ANP Topical Report EMF-2103P-A, Revision 0, Realistic Large Break Loss-of-Coolant Accident Methodology for Pressurized Water Reactors, April 9, 2003.

6.

Letter from NRC to David A Christian, Sr. Vice President and Chief Nuclear Officer, Virginia Electric Power Company, North Anna Power Station Unit 1 - Issuance of Amendment Re: Use of Framatome ANP Advanced Mark-BW Fuel, August 20, 2004.

7.

NUREG-1431, Standard Technical Specifications, Westinghouse Plants, Revision 3, TS5.6.5.

Principal Contributors: Richard Lobel Frank Orr Date: July 3, 2008