ML081440258

Reactor Vessel Surveillance Capsule Test Results for Capsule Z (Tac MD7393)

ML081440258
Person / Time
Site:	Watts Bar
Issue date:	05/28/2008
From:	Milano P NRC/NRR/ADRO/DORL/WBSPB
To:	Campbell W Tennessee Valley Authority
Milano P, NRR/DORL , 415-1457
References
TAC MD7393
Download: ML081440258 (8)

Text

May 28, 2008 Mr. William R. Campbell, Jr.

Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga Tennessee 37402-2801

SUBJECT:

WATTS BAR UNIT 1 - REACTOR VESSEL SURVEILLANCE CAPSULE TEST RESULTS FOR CAPSULE Z (TAC NO. MD7393)

Dear Mr. Campbell:

By letter dated November 9, 2007, Tennessee Valley Authority (TVA) submitted its reactor pressure vessel (RPV) capsule test results for surveillance capsule Z, which was removed from the Watts Bar Nuclear Plant Unit 1 (WBN Unit 1) RPV during the Cycle 7 refueling outage in 2007. The results were submitted to the Nuclear Regulatory Commission (NRC), pursuant to Appendix H, Reactor Vessel Material Surveillance Program Requirements, to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR). TVA provided Westinghouse Electric Company, LLC, Technical Report WCAP[Westinghouse Commercial Atomic Power]-16760-NP, Revision 0, Analysis of Capsule Z from the Tennessee Valley Authority Watts Bar Unit 1 Reactor Vessel Radiation Surveillance Program, to comply with the above requirements. The report provides the applicable fracture toughness test data for surveillance capsule Z.

The NRC staff has performed its review of report number WCAP-16760-NP, Revision 0, and has confirmed that the reports include all of the data and test results that are required by Appendix H to 10 CFR Part 50 and American Society of Testing and Materials Standard Practice E185-82, Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels. Based on its review as detailed in the enclosed safety evaluation, the NRC staff has not identified any immediate safety issues associated with the information provided in these reports. Therefore, the NRC staff intends to forward this report to the Pacific Northwest

W. Campbell, Jr. National Laboratory for the purpose of officially updating the surveillance data in the staffs Reactor Vessel Integrity Database. As required by 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, TVA shall incorporate the updated surveillance data into the next revision of the plants pressure-temperature limits.

Sincerely,

/RA/

Patrick D. Milano, Senior Project Manager Watts Bar Special Projects Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-390

Enclosure:

Safety Evaluation cc w/encl: See next page

W. Campbell, Jr. National Laboratory for the purpose of officially updating the surveillance data in the staffs Reactor Vessel Integrity Database. As required by 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, TVA shall incorporate the updated surveillance data into the next revision of the plants pressure-temperature limits.

Sincerely,

/RA/

Patrick D. Milano, Senior Project Manager Watts Bar Special Projects Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-390

Enclosure:

Safety Evaluation cc w/encl: See next page/

DISTRIBUTION:

PUBLIC RidsNrrDciCvib RidsNrrDssSrxb RidsOgcRp LWPB Reading File RidsNrrPMPMilano J. Oxendine RidsRgn2MailCenter RidsNrrDorlLwbsp RidsNrrLABClayton L. Lois RidsAcrsAcnw&mMailCenter RidsNrrDssSrxb S.Cuadrador Dejesus ADAMS Accession Number: ML081440258 OFFICE LWPB/NSPDP LWPB/PM LWPB/LA SRXB/BC CVIB/BC LWPB/BC GCranston MMitchell LRaghavan/

NAME SCuadrado PMilano BClayton by memo dated by memo dated JW for DATE 05 / 27 /08 05 / 27 /08 05 / 27 /08 02/12/08 03/24/08 05 / 28 /08 OFFICIAL RECORD COPY

William R. Campbell, Jr.

Tennessee Valley Authority WATTS BAR NUCLEAR PLANT cc:

Mr. Gordon P. Arent Mr. Michael A. Purcell New Generation Licensing Manager Senior Licensing Manager Tennessee Valley Authority Nuclear Power Group 5A Lookout Place Tennessee Valley Authority 1101 Market Street 4X Blue Ridge Chattanooga, TN 37402-2801 1101 Market Street Chattanooga, TN 37402-2801 Mr. Ashok S. Bhatnagar Senior Vice President Ms. Beth A. Wetzel, Manager Nuclear Generation Development Corporate Nuclear Licensing and and Construction Industry Affairs Tennessee Valley Authority Tennessee Valley Authority 6A Lookout Place 4X Blue Ridge 1101 Market Street 1101 Market Street Chattanooga, TN 37402-2801 Chattanooga, TN 37402-2801 Vice President Mr. Masoud Bajestani, Vice President Nuclear Support Watts Bar Unit 2 Tennessee Valley Authority Watts Bar Nuclear Plant 3R Lookout Place Tennessee Valley Authority 1101 Market Street P.O. Box 2000 Chattanooga, TN 37402-2801 Spring City, TN 37381 Mr. H. Rick Rogers Mr. Michael K. Brandon, Manager Vice President Licensing and Industry Affairs Nuclear Engineering & Technical Services Watts Bar Nuclear Plant Tennessee Valley Authority Tennessee Valley Authority 3R Lookout Place P.O. Box 2000 1101 Market Street Spring City, TN 37381 Chattanooga, TN 37402-2801 Mr. Michael J. Lorek, Plant Manager General Counsel Watts Bar Nuclear Plant Tennessee Valley Authority Tennessee Valley Authority 6A West Tower P.O. Box 2000 400 West Summit Hill Drive Spring City, TN 37381 Knoxville, TN 37902 Senior Resident Inspector Mr. John C. Fornicola, Manager Watts Bar Nuclear Plant Nuclear Assurance U.S. Nuclear Regulatory Commission Tennessee Valley Authority 1260 Nuclear Plant Road 3R Lookout Place Spring City, TN 37381 1101 Market Street Chattanooga, TN 37402-2801 County Executive 375 Church Street Mr. Larry E. Nicholson, General Manager Suite 215 Performance Improvement Dayton, TN 37321 Tennessee Valley Authority 3R Lookout Place County Mayor 1101 Market Street P. O. Box 156 Chattanooga, TN 37402-2801 Decatur, TN 37322 Mr. Michael D. Skaggs Mr. Lawrence E. Nanney, Director Site Vice President Division of Radiological Health Watts Bar Nuclear Plant Dept. of Environment & Conservation Tennessee Valley Authority Third Floor, L and C Annex P. O. Box 2000 401 Church Street Spring City, TN 37381 Nashville, TN 37243-1532

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REGARDING REACTOR VESSEL SURVEILLANCE CAPSULE TEST RESULTS FOR CAPSULE Z WATTS BAR NUCLEAR PLANT, UNIT 1 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-390

1.0 INTRODUCTION

By letter dated November 9, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML073200243), Tennessee Valley Authority (TVA, the licensee),

submitted, pursuant to Appendix H, Reactor Vessel Material Surveillance Program Requirements, to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR), its reactor pressure vessel (RPV) capsule test results for Watts Bar Nuclear Plant Unit, 1 (WBN Unit 1). In this regard, the licensee removed surveillance capsule Z from the WBN Unit 1 RPV during the Cycle 7 refueling outage in 2007 after 9.37 effective full-power years (EFPY) of operation. The licensee provided Westinghouse Electric Company, LLC, Technical Report WCAP

[Westinghouse Commercial Atomic Power]-16760-NP, Revision 0, Analysis of Capsule Z from the Tennessee Valley Authority Watts Bar Unit 1 Reactor Vessel Radiation Surveillance Program, to comply with the above requirements. The report provides the applicable fracture toughness test data for surveillance capsule Z.

2.0 REGULATORY EVALUATION

The NRC has promulgated regulations that ensure the structural integrity of the RPVs for light-water-cooled power reactors. Specific fracture toughness requirements for normal operation and for anticipated operational occurrences for power reactors are set forth in Appendix G to 10 CFR Part 50. The requirements of Appendix G are imposed by 10 CFR 50.60, Acceptance criteria for fracture prevention measures for light-water nuclear power reactors for normal operation. Additionally, in response to concerns over pressurized thermal shock events in pressurized-water reactors, the NRC issued 10 CFR 50.61. To satisfy the requirements of both Appendix G and 10 CFR 50.61, methods for determining fast neutron fluence are necessary to estimate the fracture toughness of the pressure vessel materials.

Appendix H to 10 CFR Part 50 requires the installation of surveillance capsules, including material test specimens and flux dosimeters, to provide data for material damage correlations as a function of fluence.

Appendix H provides the surveillance and testing requirements for ferritic components in the RPVs of light-water reactors. The rule requires licensees to install a number of surveillance capsules within the cavities of the RPVs and to remove capsules and test the capsule materials Enclosure

in accordance with the withdrawal schedule and testing requirements of American Society for Testing and Materials (ASTM) Standard Practice E185-82, Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels. Paragraph IV.A of the rule requires the RPV material surveillance capsule test results to be the subject of a summary technical report that is required to be submitted to the NRC within 1 year of the capsule withdrawal date.

Paragraph IV.B specifies that these topical reports shall include all data required by ASTM Standard Practice E185 and the results of all fracture toughness tests conducted on the RPV beltline materials in both the unirradiated and irradiated condition.

The NRC staff also considered the requirements of Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50 in its review. Specifically, General Design Criteria (GDC) 14, 30, and 31 are applicable. GDC 14, Reactor Coolant Pressure Boundary, requires the design, fabrication, erection, and testing of the reactor coolant pressure boundary so as to have an extremely low probability of abnormal leakage of rapidly propagating failure, and of gross rupture. GDC 30, Quality of Reactor Coolant Pressure Boundary, requires, among other things, that components comprising the reactor coolant pressure boundary be designed, fabricated, erected, and tested to the highest quality standards practical. GDC 31, Fracture Prevention of Reactor Coolant Pressure Boundary, pertains to the design of the reactor coolant pressure boundary, stating:

The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions, (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating maintenance, testing and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws.

NRC Regulatory Guide (RG) 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, describes methods and assumptions acceptable to the NRC staff for determining the RPV neutron fluence with respect to the regulatory requirements discussed above.

3.0 TECHNICAL EVALUATION

3.1 Neutron Fluence Determination The guidance provided in RG 1.190 indicates that the following comprises an acceptable fluence calculation:

a. A fluence calculation performed using an acceptable methodology

b. Analytic uncertainty analysis identifying possible sources of uncertainty

c. Benchmark comparison to approved results of a test facility

d. Plant-specific qualification by comparison to measured fluence values

The fast neutron exposure parameters were determined for the licensee by Westinghouse, using the methodologies discussed in WCAP-14040-NP-A, Methodology Used to Develop Cold Overpressure Mitigating Systems Setpoints and RCS Heatup and Cooldown Limit Curves. As noted by safety evaluation dated February 27, 2004 (ADAMS Accession No. ML040620297),

this report describes a methodology that the staff found acceptable.

For the neutron transport calculations, the licensee is using the two-dimensional discrete ordinates code, DORT, with the BUGLE-96 cross section library, which was derived from the Evaluated Nuclear Data File (ENDF/B-VI). Approximations include a P5 Legendre expansion for anisotropic scattering and a S16 order of angular quadrature. These approximations are of a higher order than the P3 expansion and S8 quadrature suggested in RG 1.190. Space and energy dependent core power (neutron source) distributions and associated core parameters are treated on a fuel cycle specific basis. Three dimensional flux solutions are constructed using a synthesis of azimuthal, axial, and radial flux. Source distributions include cycle-dependent fuel assembly initial enrichments, burnups, and axial power distributions, which are used to develop spatial and energy dependent core source distributions that are averaged over each fuel cycle. This method accounts for source energy spectral effects by using an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history of each fuel assembly. The neutron transport calculations, as described above, are performed in a manner consistent with the guidance set forth in RG 1.190.

The licensee performed an analytic uncertainty analysis by combining the uncertainties associated with the individual components of the transport calculations in quadrature. The calculations were compared with the benchmark measurements from the Poolside Critical Assembly simulator at the Oak Ridge National Laboratory, and with surveillance capsule and reactor cavity measurements from the H.B. Robinson power reactor benchmark experiment.

These constitute acceptable test facilities.

The licensee provided, and the NRC staff reviewed, a direct comparison against the measured sensor reaction rates from capsule Z. For all reactions, the measured-to-calculated (M/C) ratios were very close to unity; the average ratio was 0.96 with 6.68-percent standard deviation. The distribution of M/C ratios ranged from 0.77 to 1.05. Therefore, all reaction rates were calculated within 20-percent of measured values, as suggested in RG 1.190.

The NRC staff finds that the licensees fluence determination methods employed in the analysis of surveillance capsule Z followed the guidance presented in RG 1.190 and are, therefore, acceptable.

3.2 RPV Fracture Toughness The NRC staff reviewed WCAP-16760-NP and has confirmed that the report includes all of the data and test results that are required by Paragraph IV.B of Appendix H to Part 50, and by ASTM Standard Practice E185-82.

The NRC staff has performed its review of WCAP-16760-NP, Revision 0, and has confirmed that the analysis was performed as recommended by the NRC staff.

4.0 CONCLUSION

In summary, the licensee has provided fluence calculations performed using an acceptable methodology, supported by analytic uncertainty analysis and comparison to approved test facilities, along with a plant-specific comparison of measured fluence values from surveillance capsule Z. Based on these considerations, the NRC staff concludes that the licensee has followed the guidance in RG 1.190, and the neutron exposures reported in the licensee's submittal are, therefore, acceptable.

Based on its review, the NRC staff has not identified any immediate safety issues associated with the information provided in report, WCAP-16760-NP, Revision 0. Therefore, the staff intends to forward this report to the Pacific Northwest National Laboratory for the purpose of officially updating the surveillance data in the staffs Reactor Vessel Integrity Database.

The licensee will be expected to incorporate the updated surveillance data into the next revision of the plants pressure-temperature limits, as required by 10 CFR Part 50, Appendix G.

5.0 REFERENCES

1. Regulatory Guide 1.190, Calculation and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, NRC, March 2001.

2. WCAP-14040-NP-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, Westinghouse Electric Company, LLC, May 2004.

3. DOORS 3.2, One, Two-and Three Dimensional Discrete Ordinates Neutron/Photon Transport Code System, Radiation Safety Information Computational Center (RSICC)

Computer Code Collection CCC-650, April 1998.

4. BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Sections Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications, RSICC Library Collection DLC-185, March 1996.

5. Westinghouse Electric Company, LLC, Topical Report WCAP-16760-NP, Analysis of Capsule Z from the Tennessee Valley Authority, Watts Bar Unit 1 Reactor Vessel Radiation Surveillance Program, February 2006.

Principal Contributors: Lambros Lois James H. Oxendine Date: May 28, 2008