NOC-AE-14003131, Request for Additional Information Reactor Vessel Radiation Surveillance Program, TAC MF0699

Request for Additional Information Reactor Vessel Radiation Surveillance Program, TAC MF0699

ML14135A382
Person / Time
Site:	South Texas
Issue date:	04/30/2014
From:	Gerry Powell South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-14003131, TAC MF0699
Download: ML14135A382 (7)

Text

Nuclear Operating Company South Texas Project ElectricGeneratin$Station PO BY 28,9 Wadsworth Texas 77483 __VV_

April 30, 2014 NOC-AE-14003131 10 CFR 50 Appendix H STI: 33867989 File: G25 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 South Texas Project Units 2 Docket Nos. STN 50-499 Request for Additional Information Reactor Vessel Radiation Surveillance Proaram. TAC MF0699 Reactor Vessel Radiation Surveillance Proara TAC MF0699

References:

1. Letter from Marco Ruvalcaba, STPNOC, to NRC Document Control Desk, "Reactor Vessel Radiation Surveillance Program - STP Unit 2," dated February 11, 2013. (NOC-AE-13002957) (ML130530263)

2. E-mail from B. Singhal, NRC to L. Sterling, STP Nuclear Operating Company, "Request for Additional Information - Reactor Vessel Radiation Surveillance Program," dated April 8, 2014. (ML14099A011)

By Reference 1, STP Nuclear Operating Company (STPNOC) submitted a Reactor Vessel Radiation Surveillance Program letter for South Texas Project (STP) Unit 2. This letter provided surveillance results of Capsule W, which was removed from the STP Unit 2 reactor during refueling outage 2RE1 5 (October 28 - November 22, 2011). By e-mail dated April 8, 2014, Reference 2, the NRC staff has requested additional information regarding Reference 1.

STPNOC's response to the requests for additional information is provided in Enclosure 1 to this letter.

There are no commitments in this letter.

Should you have any questions regarding this letter, please contact Rafael Gonzales, STP Licensing point-of-contact, at (361) 972-4779 or me at 361-972-7566.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on ',4'/09' ,o/V Date G.T. Powell Site Vice President RJG

Enclosure:

1. STPNOC Response to Requests for Additional Information jDcs4

NOC-AE-14003131 Page 2 of 2 cc: (electronic copy)

(paper copy)

Regional Administrator, Region IV A. H. Gutterman, Esquire U. S. Nuclear Regulatory Commission Kathryn M. Sutton, Esquire 1600 East Lamar Boulevard Morgan, Lewis & Bockius, LLP Arlington, Texas 76011-4511 Balwant K. Singal John Ragan Senior Project Manager Chris O'Hara U.S. Nuclear Regulatory Commission Jim von Suskil One White Flint North (MS 8B1) NRG South Texas LP 11555 Rockville Pike Rockville, MD 20852 Senior Resident Inspector Kevin Polio U. S. Nuclear Regulatory Commission Cris Eugster P. O. Box 289, Mail Code: MN1 16 L.D. Blaylock Wadsworth, TX 77483 City Public Service C. M. Canady Peter Nemeth City of Austin Crain Caton & James, P.C.

Electric Utility Department 721 Barton Springs Road Austin, TX 78704 C. Mele City of Austin Jim Collins City of Austin Electric Utility Department Robert Free 721 Barton Springs Road Texas Department of State Health Services Austin, TX 78704 John W. Daily Richard A. Ratliff License Renewal Project Manager (Safety) Alice Rogers U.S. Nuclear Regulatory Commission Texas Department of State Health Services One White Flint North (MS 011-Fl)

Washington, DC 20555-0001 Tam Tran Balwant K. Singal License Renewal Project Manager John W. Daily (Environmental) Tam Tran U. S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission One White Flint North (MS 011 F01)

Washington, DC 20555-0001

Enclosure 1 NOC-AE-14003131 Enclosure 1 STPNOC Response to Requests for Additional Information

Enclosure 1 NOC-AE-14003131 Page 1 of 4 South Texas Project, Unit 2, Request for Additional Information Reactor Vessel Radiation Surveillance Program By letter dated February 11, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML130530263), STP Nuclear Operating Company, submitted WCAP-17636, "Analysis of Capsule W from the South Texas project Nuclear Operating Company Unit 2 Reactor Vessel Radiation Surveillance Program," October 2012, documenting the results of the examination of Capsule W. The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the report and has the following questions:

1. The licensee modeled the reactor and vessel using a three-dimensional representation of the problem geometry, as described in Chapter 6 of WCAP-1 7636-NP (Enclosure to letter dated February 11, 2013). The model used 208 radial, 197 azimuthal, and 190 axial intervals. The model is octant-symmetric. The azimuthal nodalization exceeds the 40 angular intervals suggested in Regulatory Guide (RG) 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001 (ADAMS Accession No. ML010890301). The recommendations for nodalization in the radial and axial directions are specific to the material and location being modeled. The RG notes that, "the adequacy of the spatial mesh and angular quadrature, as well as the convergence criterion, must be demonstrated by tightening the numerics until the resulting changes are negligible."

Please provide additional information explaining how the adequacy of the spatial mesh was validated.

2. The core neutron source was constructed based on fuel assembly-specific enrichment and burnup data for each fuel cycle of operation; this is consistent with the guidance contained in RG 1.190. The fuel assembly-specific neutron sources were derived from pin-wise isotopics represented in Cartesian geometry. The isotopic data were converted to cylindrical mesh arrays used in the RAPTOR-M3G transport calculation.

Please explain how the uncertainty associated with this conversion was determined and incorporated into the analytic uncertainty analysis, and provide additional supporting detail (i.e., provide a more detailed description of the H. B. Robinson qualification effort).

3. Section 6.1 of WCAP-17636-NP includes the statement, "Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-1 4040-A, Revision 4, 'Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves." Noting this statement, the NRC staff observes that the use of RAPTOR-M3G is a significant departure from the approved methodology in that the transport calculations are no longer performed using the suite of Oak Ridge National Laboratory discrete ordinates radiation transport codes, typically with a synthesis of lower-dimension calculations to determine the 3-dimensional flux field.

Please explain, more specifically, how the calculations are consistent with the WCAP-14040-A, Revision 4, methods. Please also identify any other differences between approved methodology and that described in WCAP-1 7636-NP.

Enclosure 1 NOC-AE-14003131 Page 2 of 4 NRC RAI Request 1:

The licensee modeled the reactor and vessel using a three-dimensional representation of the problem geometry, as described in Chapter 6 of WCAP-17636-NP (Enclosure to letter dated February 11, 2013). The model used 208 radial, 197 azimuthal, and 190 axial intervals. The model is octant-symmetric. The azimuthal nodalization exceeds the 40 angular intervals suggested in Regulatory Guide (RG) 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001 (ADAMS Accession No. ML010890301). The recommendations for nodalization in the radial and axial directions are specific to the material and location being modeled. The RG notes that, "the adequacy of the spatial mesh and angular quadrature, as well as the convergence criterion, must be demonstrated by tightening the numerics until the resulting changes are negligible."

Please provide additional information explaining how the adequacy of the spatial mesh was validated.

STPNOC RAI Request 1 Response:

When constructing the three-dimensional (3-D) RAPTOR-M3G transport model, special attention has been paid to vary mesh sizes with material and region geometry strictly in line with the guidance of Regulatory Guide 1.190 (RG 1.190). Specifically, the radial mesh in the peripheral assemblies is finer than -2 intervals per inch as required by RG 1.190 (mesh size equal to or less than 1 cm per mesh in periphery assembly region). In the excore regions, again a much finer mesh (5 to 10 intervals per inch in water, and more than 3 intervals per inch in stainless steel) is used, which is significantly finer than the -3 intervals per inch in water and -1.5 intervals per inch in steel as required by RG 1.190. Finer spatial meshes are also used in regions exhibiting steep gradients, in materials with small mean free paths, and at material interfaces, to ensure no flux changes in any group will be greater than factor 2 between adjacent intervals. For example, a very fine spatial mesh (in both r- and 0-directions) is used in modeling the stainless steel baffle plates at the periphery of the core to adequately describe this rectilinear component in r-0-z geometry (Refer to Figures 6-1 through 6-3 in WCAP-1 7636). Mesh size is selected to ensure that the flux in each spatial mesh for all energy groups reaches convergence criterion of 0.001.

In Section 1.3.1 of RG 1.190, it is stated "An azimuthal (0) mesh using at least 40 intervals over an octant in (r, 0) geometry in the horizontal plane should provide an accurate representation of the spatial distribution of the material compositions and source described in Regulatory Position 1.2." The 3-D RAPTOR-M3G model built for South Texas Unit 2 is comprised of 197 azimuthal mesh, which exceeds the requirements of RG 1.190. In general, the finer the mesh, the results are more reliable. The geometrical convergence, i.e., the impact on the discrete ordinate calculation results due to refining the spatial mesh size, should have been achieved in the RAPTOR-M3G model since it used finer mesh sizes than those required by RG 1.190.

Finally, the comparison of the RAPTOR-M3G results with the measurement data presented in Appendix A in WCAP-1 7636-NP also shows good agreement, further validating the results from 3-D RAPTOR-M3G transport calculation are credible.

Enclosure 1 NOC-AE-14003131 Page 3 of 4 NRC RAI Request 2:

The core neutron source was constructed based on fuel assembly-specific enrichment and burnup data for each fuel cycle of operation; this is consistent with the guidance contained in RG 1.190. The fuel assembly-specific neutron sources were derived from pin-wise isotopics represented in Cartesian geometry. The isotopic data were converted to cylindrical mesh arrays used in the RAPTOR-M3G transport calculation.

Please explain how the uncertainty associated with this conversion was determined and incorporated into the analytic uncertainty analysis, and provide additional supporting detail (i.e., provide a more detailed description of the H. B. Robinson qualification effort).

STPNOC RAI Request 2 Response:

For analyses using the (r, 0, z) coordinate system, the spatial component of the neutron source is transposed from x, y to r, 0 geometry by overlaying the mesh schematic to be used in the transport calculation on the pin-by-pin array and then computing the appropriate relative source applicable to each r,0 interval. This is the standard approach that has been approved previously by NRC in generating the transport source, whether the multi-channel synthesis approach is utilized or a 3-D (r, 0, z) transport calculation is utilized. The methodology employed for this analysis is consistent with WCAP-1 4040-A, Revision 4 (see the response to Question #3). Per WCAP-14040-A, the uncertainty of the spatial distribution of the source is 4%. This 4% includes not only the uncertainty attributes from the conversion, but also the generation of the pin-by-pin source, itself. The detailed description of the H. B. Robinson qualification effort is documented in WCAP-14040-A.

Enclosure 1 NOC-AE-14003131 Page 4 of 4 NRC RAI Request 3:

Section 6.1 of WCAP-1 7636-NP includes the statement, "Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-14040-A, Revision 4, 'Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves." Noting this statement, the NRC staff observes that the use of RAPTOR-M3G is a significant departure from the approved methodology in that the transport calculations are no longer performed using the suite of Oak Ridge National Laboratory discrete ordinates radiation transport codes, typically with a synthesis of lower-dimension calculations to determine the 3-dimensional flux field.

Please explain, more specifically, how the calculations are consistent with the WCAP-14040-A, Revision 4, methods. Please also identify any other differences between approved methodology and that described in WCAP-1 7636-NP.

STPNOC RAI Request 3 Response:

WCAP-14040-A, Revision 4, describes a calculation methodology that includes multi-dimensional discrete ordinates radiation transport calculations. The approved methodology allows neutron exposure to be derived from either a two-dimensional flux synthesis technique or a higher-order three-dimensional radiation transport calculation. The use of the RAPTOR-M3G code in place of the TORT code does not constitute a departure from the approved methodology described in WCAP-14040-A, Revision 4.

The RAPTOR-M3G code is a multi-dimensional discrete ordinates radiation transport code that adheres to the same discrete ordinates methodology as the TORT code. The significant distinction is that RAPTOR-M3G employs a parallel processing technique to obtain its answers. Parallel processing techniques allow large and detailed three-dimensional problems to be solved on a timescale conducive to production engineering applications. See the motivating discussion at the beginning of Appendix A of WCAP-16083-NP, Revision 1.

Appendix A.1 of WCAP-16083-NP, Revision 1 demonstrates RAPTOR-M3G and TORT produce answers that are sufficiently similar that they are interchangeable. The remaining sections in Appendix A provide additional comparisons of measurement data to calculations. RAPTOR-M3G has been designed from its inception as a parallel processing code, and adheres to the modern best practices of software development. It has been rigorously tested against the TORT code and benchmarked on an extensive set of academic and real-world problems.

For these reasons, the use of RAPTOR-M3G is considered to be methodologically consistent with WCAP-14040-A, Revision 4.