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{{Adams
#REDIRECT [[W3F1-2017-0027, Responses to Request for Additional Information Set 16 Regarding the License Renewal Application]]
| number = ML17122A176
| issue date = 05/02/2017
| title = Waterford, Unit 3 - Responses to Request for Additional Information Set 16 Regarding the License Renewal Application
| author name = Chisum M R
| author affiliation = Entergy Operations, Inc
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000382
| license number = NPF-038
| contact person =
| case reference number = W3F1-2017-0027
| document type = Letter, Response to Request for Additional Information (RAI)
| page count = 93
| project =
| stage = Response to RAI
}}
 
=Text=
{{#Wiki_filter:W3F1-2017-0027  
 
May 2, 2017
 
U.S. Nuclear Regulatory Commission
 
Attn: Document Control Desk Washington, DC  20555-0001
 
==SUBJECT:==
Responses to Request for Additional Information Set 16 Regarding the License Renewal Application for Waterford Steam Electric Station, Unit 3 (Waterford 3)
 
Docket No. 50-382 License No. NPF-38 
 
REFERENCES
:1. Entergy letter W3F1-2016-0012 "License Renewal Application, Waterford Steam Electric Station, Unit 3" dated March 23, 2016. 2. NRC letter to Entergy "Requests for Additional Information for the Review of the Waterford Steam Electric Station, Unit 3, License Renewal Application - Set 16" dated April 3, 2017. 
 
==Dear Sir or Madam:==
 
By letter dated March 23, 2016, Entergy Operations, Inc. (Entergy) submitted a license renewal application (Reference 1).
 
In letter dated April 3, 2017 (Reference 2), the NRC staff made a Request for Additional Information (RAI) Set 16 containing RAI 4.2.1-1a, needed to complete its review. Enclosure 1 provides the responses to the three requests comprising RAI 4.2.1-1a. Westinghouse letters LTR-REA-16-117 and LTR-REA-17-56 are provided as Enclosures 2 and 3, respectively. These letters and accompanying attachments provide the basis for the RAI responses presented in Enclosure 1. Additionally, Enclosure 4 contains editorial corrections to Table B-2, Sections 2.3.3.3 and B.1.4 presented in Reference 1.
 
There are no new regulatory commitments contained in this submittal. If you require additional information, please contact the Regulatory Assurance Manager, John Jarrell, at
 
504-739-6685.
17265 River Road Killona, LA 70057-3093 Tel 504-739-6660 Fax 504-739-6698 mchisum@entergy.com Site Vice President Waterford 3 
 
to W3F1-2017-0027
 
Page 1 of 2
 
==Background:==
 
In its letter dated February 6, 2017, the applicant responded to RAI 4.2.1-1 that addressed the basis of use of the RAPTOR-M3G code for its neutron fluence time-limited aging analysis (TLAA). In its response, the applicant stated that the use of RAPTOR-M3G for reactor vessel neutron fluence calculations has been incorporated into the WF3 current license basis by EC 68581. The applicant also indicated that the transition to RAPTOR-M3G was implemented via a 50.59 evaluation. 
 
LRA Section 4.2.1 states that "The methods used to calculate the WF3 vessel fluence satisfy the criteria set forth in Regulatory Guide (RG) 1.190 [Calculational and Dosimetry Methods for Determining Vessel Neutron Fluence, dated March 2001]."  Regulatory Guide (RG) 1.190 describes the guidance on calculational and dosimetry methods for determining reactor vessel neutron fluence. Specifically, Section 3 of RG 1.190 states that, when fluence determinations are required by the regulations, the licensee's documentation describing the determination of pressure vessel fluence must provide a complete description of the methods used to calculate and measure the neutron fluences. Section 3 of RG 1.190 also states that, in applying the methodology of this guide, the details of the application and the results should be reported as described in the section.
Section 3 of RG 1.190 further indicates that the reporting guidance applies to the following items: (a) fluence methods, (b) multigroup fluences; (c) integral fluences (including uncertainties); (d) comparisons of calculation and measurement; and (e) specific activities and average reaction rates. In addition, Table 1 of the RG summarizes the specific regulatory positions of reporting.
Issue:  The applicant's response does not provide relevant information (or references) to satisfy the reporting guidance of RG 1.190. The staff also noted that the applicant's response does not provide a summary of the reference document (i.e., EC 68581) to demonstrate that the use of the RAPTOR-M3G code is adequate for reactor vessel neutron fluence calculations. In addition, the staff needs additional information regarding the applicant's fluence calculations for reactor vessel nozzle areas.
Request: 1. Provide information regarding the RAPTOR-M3G-code method that demonstrates that the criteria in RG 1.190, Table 1 and Section 3 (i.e., fluence calculation methods, fluence measurement methods, reporting provisions and their associated items described in RG 1.190, Table 1) are met. 2. Describe the technical basis in EC 68581 for the use of RAPTOR-M3G code. to W3F1-2017-0027
 
Page 2 of 2
: 3. The following items can affect fluence calculations for reactor vessel nozzle areas. Therefore, explain how the following items were considered as part of the fluence calculational methodology: a) Axial distribution and isotopic content of uranium-235 and plutonium-239 in the fuel when calculating the fission source  b) Biological shield concrete composition c) Cavity gap between the reactor vessel and the biological shield  d) Homogenized materials above and below the active core region  e) Discretization effects on deterministic calculations (e.g., nozzle flux increases have resulted from changing from level symmetric (S16) to quadruple range (QR16)
 
quadrature)
: 1. Westinghouse letter LTR-REA-16-117 (Enclosure 2) provides detailed information regarding the RAPTOR-M3G code's compliance with RG 1.190 Regulatory Positions related to Section 1, Fluence Calculation Methods. Westinghouse letter LTR-REA-17-56 (Enclosure 3) provides detailed information regarding the RAPTOR-M3G code's compliance with RG 1.190 Regulatory Positions related to Section 2, Fluence Measurement Methods, and Section 3, Reporting Provisions, including a summary of the RAPTOR-M3G code's compliance with each applicable provision of Table 1 of RG 1.190. 2. Westinghouse letter LTR-REA-16-117 (Enclosure 2) provides the technical basis for the use of the RAPTOR-M3G code in EC 68581. 3. Westinghouse letter LTR-REA-17-56 (Enclosure 3) provides discussion of how the requested items were considered as part of the fluence calculational methodology. 
 
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 2 of 38 April 21, 2017 Executive Summary In the assessment of the state of embrittlement of reactor pressure vessels, an accurate evaluation of the neutron exposure for significant reactor pressure vessel materials is required. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", identifies procedures acceptable to the NRC staff for determining reactor pressure vessel neutron exposure (fluence).
This RAI response describes the methodology used to determine neutron fluence at Waterford Unit 3 in accordance with Regulatory Guide 1.190. Neutron fluence has traditionally been quantified with discrete ordinates radiation transport calculations.
The generically-approved Westinghouse fluence methodology, described in Section 2.2 of WCAP-14040-NP-A, Rev. 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," discusses the use of either the two-dimensional (2D) DORT code or the three-dimensional (3D) TORT code for calculating the transport of radiation from the reactor core to the reactor vessel wall and beyond.
Westinghouse developed RAPTOR-M3G in order to overcome limitations associated with the TORT code. RAPTOR-M3G is a 3D, parallel
-processing discrete ordinates radiation transport code that follows essentially the same calculational methodology as TORT. The parallel
-processing feature of RAPTOR
-M3G allows large, 3D radiation transport calculations to be divided across networks of workstations and solved simultaneously. This allows RAPTOR-M3G to perform calculations that would be prohibitively time consuming or impossible with TORT. This methodology for fluence determination was previously submitted on an application-specific basis for the Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Power Uprate. Per ADAMS Accession No. ML16081A333, the NRC staff determined that the methodology adequately addressed the criteria in Regulatory Guide 1.190, and was therefore acceptable.
This methodology was also previously applied to the South Texas Unit 2 Capsule W evaluation. Per ADAMS Accession No. ML14357A136, the NRC staff compared RAPTOR-M3G-generated fluence values to previously-generated values and determined that the comparisons support compliance with 10 CFR Part 50, Appendix H requirements. However, the following text was included in the correspondence: The results of the NRC staff's review, however, should not be construed as NRC approval of RAPTOR-M3G as a method of evaluation. Should future evaluations employ fluence methods that have not been NRC reviewed and approved, adequate justification regarding the application and qualification of those methods should be provided. RG 1.190 provides guidance for acceptable fluence methods.
The current RAI response provides such justification on an application
-specific basis for Waterford Unit 3.
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 3 of 38 April 21, 2017 Fluence Calculations with RAPTOR-M3G  Overview  Reactor vessel neutron fluence has traditionally been quantified using discrete ordinates radiation transport calculations. Codes used to perform early calculations include TWOTRAN (Reference
: 1) and DOT (Reference 2). With the limitations on computing power at the time, both TWOTRAN and DOT were only capable of analyzing one
-dimensional and two-dimensional models. In the 1980s, Oak Ridge National Laboratory developed the DORT (two
-dimensional) and TORT (three
-dimensional) codes (Reference 3), and these codes remain in widespread use today.
The methodology employed by RAPTOR-M3G is essentially the same as the methodology employed by the TORT code, with solution enhancements resulting from the last two decades of research.
RAPTOR-M3G has been designed from its inception as a parallel-processing code, and adheres to 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. The methodology used to provide neutron exposure evaluations for the reactor pressure vessel (RPV) follows the guidance provided in Regulatory Guide 1.190 (Reference 4). The use of RAPTOR-M3G satisfies all regulatory positions in Regulatory Guide 1.190 pertinent to neutron fluence calculation methods. Geometric Modeling In developing an analytical model of the Light Water Reactor (LWR) reactor geometry, nominal design dimensions are normally employed for the various structural components. In some cases, as-built dimensions are available; in those instances, the more accurate as-built data are used for model development. However, most as-built dimensions of the components in the beltline region of the reactor are not available; instead, design dimensions are used. Likewise, nominal full-power design values are used for water temperatures and, hence, coolant density in the reactor core and downcomer regions. The reactor core is treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, etc. The stainless steel former plates located between the core baffle and core barrel regions are also included in the model. Sensitivities of the analytical results to tolerances in the internals dimensions and fluctuations in water temperature are discussed and quantified later in this document. The geometric mesh description of the reactor model is normally accomplished using from 150 to 250 radial, 80 to 150 azimuthal, and 100 to 200 axial intervals depending on the overall size of the reactor and on the complexity required to model the core periphery, the in-vessel surveillance capsules, and the details of the reactor cavity. Mesh sizes are chosen to assure that proper convergence of the inner iterations is achieved on a pointwise basis. The pointwise inner iteration convergence criterion utilized in the transport calculations is set at a value of 0.001.
The Waterford Unit 3 model uses 160 radial mesh intervals, 121 azimuthal mesh intervals, and 247 axial mesh intervals. The peripheral assemblies and water regions were modeled with approximately 4 intervals per inch, and the steel regions were modeled with approximately 1.7 intervals per inch. Nominal design Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 4 of 38 April 21, 2017 dimensions were used to model steel components, as pertinent as-built information was not readily available for Waterford Unit 3. The mesh selection process results in a smaller spatial mesh in regions exhibiting steep gradients, in material zones of high cross section (), and at material interfaces. In the modeling of in
-vessel surveillance capsules, a sufficiently fine mesh grid is employed within the test specimen array to assure that accurate information is produced for use in the assessment of fluence gradients within the materials test specimens, as well as in the determination of gradient corrections for neutron sensors. Additional radial and azimuthal mesh are employed to model the capsule structure surrounding the materials test specimen array. In modeling the stainless steel baffle region at the periphery of the core, a relatively fine spatial mesh is required to adequately describe this rectilinear component in cylindrical geometry. In performing this Cartesian to cylindrical transition, care is taken to preserve both the thickness and volume of the steel region in order to accurately address the shielding effect of the steel. The geometric modeling described in this section and applied for Waterford Unit 3 complies with Regulatory Position 1.1.1 of Regulatory Guide 1.190.
Cross Sections and Materials The transport calculations are carried out using the BUGLE-96 cross-section library (Reference 5). The BUG LE-96 library provides a 67-group coupled neutron-gamma ray cross-section data set produced specifically for LWR application. Anisotropic scattering is treated with a minimum P 3 Legendre expansion. For the Waterford Unit 3 calculations, a (more detailed) P 5 Legendre expansion was used. Number densities used in the representation of material mixtures for LWR applications are taken from the documentation provided with BUGLE-96 for carbon steel, stainless steel, concrete, zircalloy-4, and uranium dioxide. Other mixtures, such as Inconel-718, come from internal Westinghouse data sources. The BUGLE-B7 cross-section library (Reference
: 6) is an update to the BUGLE-96 library. The primary difference between the two libraries is that the BUGLE-B7 library is derived from ENDF-B/VII.0 nuclear data, whereas BUGLE-96 is derived from older ENDF/B-VI nuclear data. The energy group boundaries and techniques used to construct both libraries are the same. In the documentation released with BUGLE-B7, Oak Ridge National Laboratory (ORNL) analyzed the H. B. Robinson Unit 2 , Pool Critical Assembly (PCA), and VENUS-3 benchmarks using BUGLE-96 and BUGLE-B7. Calculations with both libraries were compared to measurements from reactions that cover high-energy portions of the neutron spectrum that are of greatest concern for reactor vessel integrit y evaluations. The ORNL report demonstrates that the differences between BUGLE-96 and BUGLE-B7 for high-energy neutron applications are minor. This finding is corroborated by comparisons that Westinghouse has performed internally. Thus, for applications that concern Regulatory Guide 1.190, the differences between BUGLE-96 and BUGLE-B7 are not significant.
Westinghouse has completed an additional comparative study that revealed differences between BUGLE-B7 and BUGLE-96 for low energy neutrons that resulted from discrepancies in the upscatter
-removed BUGLE-B7 library in the lower energy range. This discrepancy does not affect the validity of Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 5 of 38 April 21, 2017 the BUGLE-B7 library for reactor vessel integrity evaluations, but Westinghouse has chosen to continue using BUGLE-96 until the discrepancy with BUGLE
-B7 is resolved.
The use of the BUGLE
-96 cross-section library with the P 5 Legendre expansion mode for Waterford Unit 3 and comparisons to the newer BUGLE-B7 library described in this section complies with Regulatory Position 1.1.2 of Regulatory Guide 1.190. Core Source Definition The spatial variation of the neutron source is obtained from a burnup weighted average of the respective power distributions from individual fuel cycles. These spatial distributions include pinwise gradients for all fuel assemblies located at the periphery of the core and include a uniform or flat distribution for fuel assemblies interior to the core. The spatial component of the neutron source is transposed from Cartesian to cylindrical 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 cylindrical mesh interval. The Cartesian
-to-cylindrical transposition is accomplished by first defining a fine cylindrical mesh working array. The  and  mesh are chosen so that there is typically a 10x10 array of fine mesh over the area of each fuel pin at the core periphery. The coordinates of the center of each fine mesh interval and its associated relative source strength are assigned to the fine mesh based on the pin that is coincident with the center of the fine mesh. In the limit as  and  approach zero, this technique becomes an exact transformation.
Each space mesh in the cylindrical transport geometry is checked to determine if it lies totally within the area of a particular fine working mesh. If it does, the relative source of that fine mesh is assigned to the transport space mesh. If, otherwise, the transport space mesh covers a part of one or more fine mesh, then the relative source assigned to the transport mesh is determined by an area weighting process as follows:
=  (1) where:  = the relative source assigned to transport mesh m
  = the area of fine working mesh i within transport mesh m
  = the relative source within fine working mesh i The energy distribution of the source is determined by selecting a fuel burnup representative of conditions averaged over the irradiation period under consideration and an initial fuel assembly enrichment characteristic of the core designs used over the applicable period. From the assembly burnup and initial U-235 enrichment, a fission split by isotope including U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242 is derived; and, from that fission split, composite values of energy release per fission, neutron Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 6 of 38 April 21, 2017 yield per fission, and fission spectrum are determined. These composite values are then combined with the spatial distribution to produce the overall absolute neutron source for use in the transport calculations. The core neutron source definition methodology described in this section complies with Regulatory Position 1.2 in Regulatory Guide 1.190.
Discrete Ordinates Calculations with RAPTOR-M3G RAPTOR-M3G solves the time-independent Linear Boltzmann Equation (LBE) in the absence of fission in three dimensions via the discrete ordinates approximation. The method of discrete ordinates is described in detail in Reference
: 3. The following methodological differences exist between the RAPTOR-M3G code and the TORT code:
RAPTOR-M3G does not use zero
-weighted initiating directions.
See Page 3
-33 of the TORT manual in Reference 3 for a discussion of this difference. RAPTOR-M3G uses the technique of Lathrop and Brinkley, which has been established to provide good results. Further, the documented consistency between RAPTOR-M3G and TORT in the calculated fast neutron (E > 1.0 MeV) fluence rate responses in Reference 7 suggests that this difference is practically insignificant for reactor vessel neutron fluence calculations. In particular, note the similarity of the values listed in the bottom two rows of Table 2-2 through Table 2-4 of Reference
: 7. RAPTOR-M3G includes an implementation of the Directional Theta-Weighted (DTW) spatial differencing scheme, whereas TORT does not. The DTW scheme is discussed and endorsed in Regulatory Guide 1.190. (See Reference 34 of Regulatory Guide 1.190 for more information about the DTW scheme.) The DTW scheme generally produces improved results, as compared to traditional theta
-weighted (TW) schemes.
RAPTOR-M3G calculations are performed with an S 8 (or higher) level
-symmetric angular quadrature set.
Neutron fluence values are determined directly from the results of the radiation transport calculations performed with RAPTOR-M3G. For Waterford Unit 3, the calculations were performed in DTW mode with an S 16 level-symmetric angular quadrature set. The RAPTOR
-M3G calculations described in this section (and inputs described in previous sections) comply with Regulatory Positions 1.3.1 and 1.3.3 of Regulatory Guide 1.190. Regulatory Position 1.3.2 is not applicable because deterministic (not Monte Carlo) calculations are being performed. Regulatory Position 1.3.4 is not applicable because 3-D calculations are performed, and the synthesis technique is not used. Regulatory Position 1.3.5 is not applicable because cavity dosimetry was not analyzed.
 
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 7 of 38 April 21, 2017 Fluence Calculation Methodology Qualification for RAPTOR-M3G Overview The validation of the transport methodology in RAPTOR-M3G follows the guidance provided in Regulatory Guide 1.190. In particular, the validation consists of the following stages:
: 1. Simulator Benchmark Calculations:
comparisons of calculations with measurements from simulator benchmarks, including the PCA simulator (Reference
: 8) at ORNL and the VENUS-1 experiment (Reference 10). 2. Operating Reactor and Calculational Benchmarks:
comparisons of calculations with surveillance capsule and reactor cavity measurements from the H. B. Robinson Unit 2 power reactor benchmark experiment (Reference 12), and comparisons of calculations performed with RAPTOR-M3G to results published in the NRC's fluence calculation benchmark (Reference 13). 3. Analytic Uncertainty Analysis:
an analytic sensitivity study addressing the uncertainty components resulting from important input parameters applicable to the plant
-specific transport calculations used in the exposure assessments.
At each subsequent application of the methodology, comparisons are made with plant-specific dosimetry results to demonstrate that the plant
-specific transport calculations are consistent with the uncertainties derived from the methods qualification. The first stage of the methods validation addresses the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross sections. This phase, however, does not test the accuracy of commercial core neutron source calculations nor does it address uncertainties in operational or geometric variables that affect power reactor calculations. The second stage of the validation addresses uncertainties that are primarily methods related and would tend to apply generically to all fast neutron exposure evaluations. The third stage of the validation identifies the potential uncertainties introduced into the overall evaluation due to calculational methods approximations as well as to a lack of knowledge relative to various plant
-specific parameters. The overall calculational uncertainty is established from the results of these three stages of the validation process.
Simulator Benchmark Calculations Several simulator benchmark experiments have been performed for the purpose of providing a qualification basis for neutron fluence analysis methods. The experiments were performed in laboratory settings, and simulate the configuration of an operating nuclear reactor on a smaller scale. This section provides the results of comparisons of simulator benchmark measurement results with calculations performed with RAPTOR-M3G.
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 8 of 38 April 21, 2017 Simulator Benchmark Calculations - PCA The PCA Pressure Vessel Facility Benchmark (Reference
: 8) is an industry
-standard benchmark that can be used to partially qualify a fluence determination methodology according to Regulatory Guide 1.190. The PCA facility provides a small-scale simulation of the configuration of a Pressurized Water Reactor
 
(PWR). The geometry, material compositions, and neutron source for this experiment were all well
-characterized, and accurate dosimetry measurements were collected at several locations of interest. A complete description of the benchmark is available in Reference
: 8. Table 1 shows the distribution of measurement locations. The RAPTOR
-M3G analysis of the PCA problem with the 12/13 configuration is modeled on a 67 x 139 x 102 Cartesian mesh grid. Angular quadrature is modeled with an S 8 level-symmetric quadrature set, and anisotropic scattering is treated with a P 3 Legendre expansion. The transport cross
-section set was constructed from the BUGLE-96 library, and dosimetry reaction rate cross
-sections are taken from the SNLRML library (Reference 9). The PCA problem was analyzed in RAPTOR-M3G using the TW and DTW differencing schemes.
Results of the benchmark comparisons using RAPTOR-M3G are presented in Table 2 for TW differencing and Table 3 for DTW differencing. Table 1: PCA Experimental Measurement Locations Position Y (cm) Location Description A1 A2 A3 A4 A5 A6 A7 12.0 23.8 29.7 39.5 44.7 50.1 59.1 Thermal Shield (Front)
Thermal Shield (Back)
Pressure Vessel (Front)
Pressure Vessel (1/4 T)
Pressure Vessel (1/2 T)
Pressure Vessel (3/4 T)
Void Box Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 9 of 38 April 21, 2017 Table 2: Measurement
-to-Calculation (M/C) Reaction Rate Comparisons  for the PCA 12/13 Blind Test Experiment (TW Differencing)
Reaction M/C Ratio for Dosimetry Position Noted A1 A2 A3 A4 A5 A6 A7 27Al (n,) 24Na (Cd) 0.99 1.00 0.95 0.95 0.96 0.98 - 58Ni (n,p) 58Co (Cd) 1.03 1.03 0.99 1.00 1.01 0.97 - 115115mIn (Cd) 1.03 1.03 0.97 0.97 0.99 1.00 1.06 103103mRh (Cd) 1.01 0.98 0.98 0.98 1.04 1.06 1.07 238 U (n,f) FP (Cd)
- - 0.92 1.01 1.03 1.06 1.06 237Np (n,f) FP (Cd)
- - 1.07 1.00 0.99 1.03 0.98 Average 1.02 1.01 0.98 0.99 1.00 1.02 1.04 % std dev 1.9 2.4 5.2 2.3 2.9 3.9 4.0    Table 3: M/C Reaction Rate Comparisons  for the PCA 12/13 Blind Test Experiment (DTW Differencing)
Reaction M/C Ratio for Dosimetry Position Noted A1 A2 A3 A4 A5 A6 A7 27Al (n,) 24Na (Cd) 0.98 0.99 0.95 0.97 0.99 1.00 - 58Ni (n,p) 58Co (Cd) 1.02 1.02 0.99 1.01 1.02 0.98 - 115115mIn (Cd) 1.02 1.02 0.96 0.97 1.00 1.00 1.06 103103mRh (Cd) 1.00 0.97 0.98 0.98 1.04 1.07 1.08 238U (n,f) FP (Cd)
- - 0.92 1.01 1.04 1.06 1.07 237Np (n,f) FP (Cd)
- - 1.06 1.01 1.00 1.03 0.99 Average 1.01 1.00 0.98 0.99 1.02 1.02 1.05 % std dev 1.9 2.4 4.9 2.1 2.1 3.5 3.9 
 
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 10 of 38 April 21, 2017 Simulator Benchmark Calculations - VENUS-1 Benchmark The VENUS-1 experiment (Reference
: 10) is another commonly-used qualification benchmark. As with the PCA benchmark, the critical variables affecting the measurements were carefully measured and recorded. The VENUS-1 benchmark correctly represents the heterogeneities in a PWR, and includes a stainless steel core baffle, core barrel, and neutron pad. The benchmark experiment was performed at room temperature (300 K). Forty-one measurement locations exist in the benchmark, which are given in Table 4. The RAPTOR
-M3G analysis of the VENUS-1 problem is modeled on a 192 x 123 x 65 cylindrical mesh grid. Angular quadrature is modeled with an S 8 level-symmetric quadrature set, and anisotropic scattering is treated with a P 3 Legendre expansion. The transport cross
-section set was constructed from the BUGLE-96 library, and dosimetry reaction rate cross sections are taken from the SNLRML library.
Results of the benchmark comparisons using RAPTOR-M3G are presented in Table 5 for TW differencing and Table 6 for DTW differencing.
Table 4: VENUS-1 Measurement Locations Point Description 1 Central Water Hole 2-3 Inner Baffle 4-10 Outer Baffle 11-14 Along 45° in 3/0 fuel region 15-19 Along 45° in Water Gap I 20-27 Core Barrel (0-45°) 28-36 At a radius of 55.255 cm in Water Gap II 37-41 Neutron Pad
 
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 11 of 38 April 21, 2017 Table 5: M/C Reaction Rate Comparisons for the VENUS
-1 Experiment (TW Differencing)
Point M/C Ratio 58Ni (n,p) 115In (n,n') 103Rh (n,n') 238U (n,f) 237Np (n,f) 1 1.01 0.99 0.97 1.11 - 2 0.97 0.97 0.93 1.03 0.98 3 0.96 0.97 0.93 1.02 0.96 4 0.98 0.97 - 1.05 1.06 5 0.98 0.96 0.94 1.02 - 6 0.99 0.96 0.93 1.04 1.03 7 1.01 0.97 0.94 - 1.02 8 1.03 0.97 0.94 1.03 1.03 9 0.96 0.96 0.93 1.04 1.00 10 0.97 0.98 0.95 1.01 1.01 11 - - - - 0.98 12 - - - - 1.00 13 - - - - 0.98 14 - - - - 1.08 15 - 0.98 0.97 1.04 0.98 0.97 - 0.98 17 - 0.98 0.99 1.06 0.99 18 - 1.00 0.93 - 1.02 19 - 0.99 1.01 1.08 - 20 0.97 0.99 - 1.06 0.98 21 0.99 0.98 - 1.06 0.99 22 1.00 0.98 0.93 1.08 1.07 23 1.03 0.99 - 1.07 1.07 24 1.05 0.98 - 1.10 1.05 25 1.07 0.98 - 1.11 0.99 26 1.07 0.99 - 1.05 1.06 27 1.08 0.99 0.98 1.07 1.05 28 - 1.03 - 1.13 1.07 29 - 1.01 - 1.08 1.10 30 - 1.02 - 1.16 1.02 - 1.08 1.06 32 - 1.04 - 1.10 1.11 33 - 1.01 - 1.10 1.05 34 - - - 1.12 1.11 35 - 1.08 - - 1.00 - 1.09 1.13 37 - - - - 1.12 - - - - - - - - 1.09 - - -
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 12 of 38 April 21, 2017 Table 6: M/C Reaction Rate Comparisons for the VENUS-1 Experiment (DTW Differencing)
Point M/C Ratio 58Ni (n,p) 115In (n,n') 103Rh (n,n') 238U (n,f) 237Np (n,f) 1 1.00 0.98 0.95 1.10 - 2 0.98 0.98 0.94 1.04 0.99 3 0.97 0.98 0.93 1.03 0.97 4 0.98 0.98 - 1.05 1.06 5 0.98 0.97 0.95 1.03 - 6 1.00 0.97 0.94 1.05 1.03 7 1.03 0.98 0.95 - 1.04 8 1.03 0.98 0.95 1.04 1.04 9 0.96 0.97 0.93 1.04 1.01 10 0.97 0.97 0.95 1.01 1.01 11 - - - - 0.98 12 - - - - 1.00 13 - - - - 0.99 14 - - - - 1.08 15 - 0.98 0.97 1.04 0.98 0.97 - 0.99 17 - 0.99 1.00 1.07 0.99 18 - 1.01 0.94 - 1.03 19 - 1.00 1.02 1.09 - 20 0.99 1.00 - 1.08 0.99 21 1.01 1.00 - 1.08 1.01 22 1.01 1.00 0.95 1.10 1.09 23 1.05 1.01 - 1.08 1.09 24 1.07 1.00 - 1.12 1.07 25 1.08 0.99 - 1.13 1.01 26 1.09 1.02 - 1.07 1.08 27 1.11 1.01 1.00 1.09 1.08 28 - 1.04 - 1.14 1.08 29 - 1.02 - 1.10 1.11 30 - 1.03 - 1.17 1.03 - 1.09 1.07 32 - 1.05 - 1.12 1.12 33 - 1.02 - 1.11 1.06 34 - - - 1.14 1.12 35 - 1.09 - - 1.01 - 1.11 1.14 37 - - - - 1.15 - - - - - - - - 1.11 - - -
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 13 of 38 April 21, 2017 Simulator Benchmark Calculations - Summary Results of the PCA and VENUS-1 simulator benchmarks, grouped by reaction, are summarized in Table 7 for the TW differencing scheme and Table 8 for the DTW differencing scheme. Also included in both tables are the energy response ranges between which 90% of activity is produced in a U-235 fission spectrum, taken from ASTM E844 (Reference 11). The U-238 and Np-237 measurements exhibit slightly worse agreement with the calculations; however, the U-238 and Np-237 reactions are subject to higher uncertainties in the measurement process.
The simulator benchmarks test the adequacy of the transport and dosimetry evaluation techniques, and the underlying nuclear data. The simulator benchmark comparison results demonstrate that, when the configuration of the system is well
-known, the level of agreement between RAPTOR
-M3G calculations and measurements is within the uncertainties associated with the measurements, themselves. The uncertainty assigned to the calculational methodology from simulator benchmarks is 3%. This simulator benchmark comparisons contribute to addressing Regulatory Position 1.4.2 of Regulatory Guide 1.190.
 
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 14 of 38 April 21, 2017 Table 7: Summary of Simulator Benchmark M/C Reaction Rate Comparisons (TW Differencing)
Reaction Neutron Energy Response Number of Observations Average M/C
% std dev 27Al (n,) 24Na (Cd) 6.45 - 11.9 MeV 6 0.97 2.2 58Ni (n,p) 58Co (Cd) 1.98 - 7.51 MeV 24 1.01 3.6 115115mIn (Cd) 1.12 - 5.86 MeV 40 1.00 3.7 103103mRh (Cd) 0.731 - 5.73 MeV 23 0.97 4.4 238U (n,f) FP (Cd) 1.44 - 6.69 MeV 33 1.06 4.2 237Np (n,f) FP (Cd) 0.684 - 5.61 MeV 35 1.03 4.4 Total  161 1.02 5.0  Table 8: Summary of Simulator Benchmark M/C Reaction Rate Comparisons (DTW Differencing)
Reaction Neutron Energy Response Number of Observations Average M/C
% std dev 27Al (n,) 24Na (Cd) 6.45 - 11.9 MeV 6 0.98 1.8 58Ni (n,p) 58Co (Cd) 1.98 - 7.51 MeV 24 1.01 4.0 115115mIn (Cd) 1.12 - 5.86 MeV 40 1.01 3.9 103103mRh (Cd) 0.731 - 5.73 MeV 23 0.98 4.4 238U (n,f) FP (Cd) 1.44 - 6.69 MeV 33 1.07 4.5 237Np (n,f) FP (Cd) 0.684 - 5.61 MeV 35 1.04 4.5 Total  161 1.02 5.2 Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 15 of 38 April 21, 2017 Operating Reactor and Calculational Benchmarks In addition to measurements from laboratory-scale simulator benchmark experiments, Regulatory Guide 1.190 recommends that methods qualification should be based on comparisons with measurement data from operating power reactors and comparisons with reference results from calculational benchmark problems. This section provides the comparisons of power reactor measurements to calculations performed with RAPTOR
-M3G, and details the results of one calculational study. Operating Reactor and Calculational Benchmarks - H. B. Robinson Unit 2 Cycle 9 Benchmark H. B. Robinson Unit 2 is a Westinghouse 3
-loop PWR. As part of the NRC-sponsored LWR Pressure Vessel Surveillance Dosimetry Improvement Program, a comprehensive set of surveillance capsule and ex-vessel neutron dosimetry measurements were performed during Cycle 9 (Reference 12). For the Cycle 9 benchmark, a replacement surveillance capsule was installed in a vacant surveillance capsule holder at the 20° azimuth with respect to the nearest cardinal axis. The dosimetry sets were placed at the geometric center of the surveillance capsule such that all measurements were taken within 30 cm of the core midplane. The ex-vessel neutron dosimetry was installed in the reactor cavity, between the concrete biological shield and the reactor vessel insulation. Multiple foil sensor sets were placed in capsules that were attached to gradient wires. The gradient wires were installed in the reactor cavity and axially spanned the length of the reactor core. Only the midplane capsule from the Cycle 9 ex
-vessel neutron dosimetry set was analyzed. H. B. Robinson Unit 2 was modeled in cylindrical geometry with 158x136x172 mesh using RAPTOR-M3G. Angular quadrature was modeled with an S 8 level-symmetric quadrature set, and anisotropic scattering was treated with a P 3 Legendre expansion. Detailed geometry data for H. B. Robinson Unit 2 can be found in Reference
: 12. Neutron spectra obtained from the transport calculations were combined with the dosimetry cross sections from the SNLRML library (Reference
: 9) in order to obtain calculated reaction rates for comparison to the measurements. These comparisons are presented in Table 9 and Table 10 for transport results obtained with TW and DTW differencing schemes, respectively.
 
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 16 of 38 April 21, 2017 Table 9: Results for the H. B. Robinson Unit 2  Cycle 9 Dosimetry Benchmark Experiment (TW Differencing)
In-Vessel Reaction Reaction Rate [rps/atom]
M/C Ratio Measured Calculated Cu-63 (n,Co-60 3.86E-17 3.67E-17 1.05 Ti-46 (n,p) Sc
-46 6.92E-16 5.83E-16 1.19 Fe-54 (n,p) Mn-54 3.81E-15 3.67E-15 1.04 Ni-58 (n,p) Co-58 5.34E-15 4.97E-15 1.07 U-238 (n,f) FP 1.74E-14 1.64E-14 1.06 Np-237 (n,f) FP 1.19E-13 1.14E-13 1.04 Average  1.08  Ex-Vessel Reaction Reaction Rate [rps/atom]
M/C Ratio Measured Calculated Cu-63 (n,Co-60 3.86E-19 4.05E-19 0.95 Ti-46 (n,p) Sc
-46 6.55E-18 5.81E-18 1.13 Fe-54 (n,p) Mn-54 3.55E-17 3.76E-17 0.94 Ni-58 (n,p) Co-58 5.85E-17 5.65E-17 1.04 U-238 (n,f) FP 2.74E-16 2.60E-16 1.05 Average  1.02 Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 17 of 38 April 21, 2017 Table 10: Results for the H. B. Robinson Unit 2  Cycle 9 Dosimetry Benchmark Experiment (DTW Differencing)
In-Vessel Reaction Reaction Rate [rps/atom]
M/C Ratio Measured Calculated Cu-63 (n,Co-60 3.86E-17 3.68E-17 1.05 Ti-46 (n,p) Sc
-46 6.92E-16 5.84E-16 1.18 Fe-54 (n,p) Mn-54 3.81E-15 3.68E-15 1.04 Ni-58 (n,p) Co-58 5.34E-15 4.99E-15 1.07 U-238 (n,f) FP 1.74E-14 1.65E-14 1.05 Np-237 (n,f) FP 1.19E-13 1.16E-13 1.03 Average  1.07  Ex-Vessel  Reaction Reaction Rate [rps/atom]
M/C Ratio Measured Calculated Cu-63 (n,Co-60 3.86E-19 4.10E-19 0.94 Ti-46 (n,p) Sc
-46 6.55E-18 5.87E-18 1.12 Fe-54 (n,p) Mn-54 3.55E-17 3.78E-17 0.94 Ni-58 (n,p) Co-58 5.85E-17 5.71E-17 1.02 U-238 (n,f) FP 2.74E-16 2.66E-16 1.03 Average  1.01 Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 18 of 38 April 21, 2017 Operating Reactor and Calculational Benchmarks - NRC Fluence Calculation Benchmark The NRC's fluence benchmark (Reference
: 13) is a calculational exercise, developed by Brookhaven National Laboratory at the request of the NRC, which provides reference solutions for typical PWR and Boiling Water Reactor (BWR) pressure vessel fluence calculations.
The PWR problem models a typical 204 fuel assembly PWR core including the core baffle and barrel, thermal shield, and a pressure vessel. Three types of fuel loadings are defined, including a standard out-in core loading, a low-leakage core loading, and a core that includes partial length shield assemblies (PLSAs). Reference solutions derived from multigroup synthesis and Monte Carlo simulations are provided in Reference
: 13. The PWR problem was analyzed with RAPTOR-M3G. The PWR model analysis presented herein considers the standard core loading configuration. Angular quadrature is modeled with an S 8 level-symmetric quadrature set, and anisotropic scattering is treated with a P 3 Legendre expansion. Comparisons relative to the fast (E > 1.0 MeV) fluence rate synthesis solutions published in Reference 13 are provided in Table 11.
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 19 of 38 April 21, 2017 Table 11: NRC Fluence Benchmark Comparisons with RAPTOR-M3G PWR Problem with Standard Core Loading  Pressure Vessel Inner-Wall Lower Weld Location (R=219.393 cm, Z=67.1048 cm) Theta (Degrees) Fast (E > 1.0 MeV) Neutron Fluence Rate
[n/cm 2-s] RAPTOR-M3G with DTW  Relative Diff. (%) RAPTOR-M3G with TW  Relative Diff.  (%) Reference 13 , Table 4.1.2.4 RAPTOR-M3G with DTW RAPTOR-M3G with TW 1.125 2.71E+10 2.57E+10 2.55E+10 -4.99% -5.86% 3.375 2.76E+10 2.64E+10 2.63E+10 -4.48% -4.96% 5.625 2.87E+10 2.77E+10 2.75E+10 -3.49% -4.38% 7.875 3.04E+10 2.96E+10 2.94E+10 -2.78% -3.36% 10.125 3.31E+10 3.18E+10 3.16E+10 -4.03% -4.59% 12.375 3.49E+10 3.39E+10 3.38E+10 -3.01% -3.11% 14.625 3.62E+10 3.53E+10 3.52E+10 -2.44% -2.81% 16.875 3.64E+10 3.55E+10 3.56E+10 -2.29% -2.07% 19.125 3.45E+10 3.35E+10 3.32E+10 -2.79% -3.92% 21.375 3.18E+10 3.15E+10 3.11E+10 -0.97% -2.24% 23.625 2.95E+10 2.90E+10 2.87E+10 -1.75% -2.94% 25.875 2.79E+10 2.73E+10 2.67E+10 -2.48% -4.56% 28.125 2.74E+10 2.68E+10 2.64E+10 -2.04% -3.65% 30.375 2.77E+10 2.72E+10 2.68E+10 -1.95% -3.22% 32.625 2.77E+10 2.71E+10 2.72E+10 -2.01% -1.68% 34.875 2.67E+10 2.56E+10 2.57E+10 -4.00% -3.88% 37.125 2.36E+10 2.35E+10 2.34E+10 -0.77% -1.18% 39.375 2.08E+10 2.07E+10 2.06E+10 -0.59% -1.10% 41.625 1.88E+10 1.80E+10 1.77E+10 -4.31% -5.85% 43.875 1.72E+10 1.66E+10 1.62E+10 -3.70% -5.93%
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 20 of 38 April 21, 2017 Operating Reactor and Calculational Benchmarks - Summary The H. B. Robinson Unit 2 benchmark represents an experimental configuration that is broadly reflective of most operating reactors: data was collected during full
-power operation at a commercial LWR; the power distribution and power history data supporting the analysis were derived using methods similar to those employed by most operating LWRs; geometric dimensions specified are nominal dimensions, and not necessarily identical to their as-built configuration. These characteristics make the H. B. Robinson Unit 2 benchmark a compelling data set. For the H. B. Robinsion Unit 2 benchmark, the maximum average of the in-vessel M/C ratios is 1.08, and the maximum average of the ex
-vessel M/C ratios is 1.02. The uncertainty assigned to the calculational methodology from H. B. Robinson Unit 2 benchmark is 8%. The calculational benchmark do es not provide real measurement data, and the methods and data used in the reference results are somewhat dated by contemporary standards.
Therefore, results of these evaluations are not used as a direct input to the overall bias and uncertainty assessment for the fluence determination methodology. Nonetheless, the consistency of the RAPTOR-M3G results, both with the reference calculations provided by Brookhaven and the self-consistency demonstrated by the two differencing schemes in RAPTOR-M3G, provides additional confidence that RAPTOR-M3G is correctly applying the discrete ordinates method. This operating reactor benchmark comparisons and comparisons to the calculational benchmark contribute to addressing Regulatory Position 1.4.2 of Regulatory Guide 1.190.
 
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 21 of 38 April 21, 2017 Analytic Uncertainty Analysis Operating reactors are subject to several uncertainties that may influence the validity of the calculated neutron fluence results. The most significant among these are:
Uncertainties in the core neutron source Uncertainties in the as
-built thicknesses and locations of the reactor vessel and internal components  Uncertainties in the full-power coolant temperatures (water density)
This listing of parameters is consistent with the findings of other neutron fluence uncertainty studies (References 14 , 15, and 16). This section presents the results of a sensitivity study performed using RAPTOR-M3G that evaluate the impacts of variations in the parameters listed above on calculated neutron fluence values.
Note that the uncertainty analysis was performed for both the TW and DTW differencing schemes. In general, the analytic uncertainty values are consistent between the two differencing schemes; however, in cases where there are differences, the higher uncertainty values were selected.
Analytic Uncertainty Analysis - Core Neutron Source Uncertainties To assess the impact of uncertainties in the core neutron source on calculated neutron fluence results, changes in the following parameters were evaluated:
Absolute source strength of peripheral fuel assemblies - Studies have shown that the neutron fluence rate in regions external to the core is dominated by the neutron source from fuel assemblies on the core periphery. In-core measurements indicate that a source magnitude uncertainty of 5% is bounding.
Pin-by-pin spatial distributions of neutron source at the core periphery - Core management studies indicate that uncertainties in the relative pin powers in peripheral fuel assemblies can be on the order of 10%. Burnup of the peripheral fuel assemblies - Perturbations in fuel assembly burnup impact the fission spectrum, neutron yield per fission, and energy released per fission for each peripheral fuel assembly. A 5000 MWD/MTU uncertainty in the peripheral fuel assembly burnups is considered conservative. The sensitivity study is performed using a series of calculations starting with mid
-cycle burnup at 3000 MWD/MTU, and 5000 MWD/MTU to 50,000 MWD/MTU with 5000 MWD/MTU delta mid-cycle burnup between each run. Axial power distribution - Based on variations in axial peaking factors over the course of a fuel cycle, a 10% uncertainty in the shape of the axial power distribution is considered conservativ
: e.
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 22 of 38 April 21, 2017 Each case evaluated as part of the sensitivity study is described in Table 12. The base case consisted of a low-leakage power distribution cycle from a Westinghouse 4-Loop reactor. Table 13 through Table 15 provides the differences between calculated fast neutron (E >
1.0 MeV) fluence rate results at several locations for each permutation case, each normalized to the corresponding base case result. The overall uncertainty estimates are summarized in Table 16 through Table 18.
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 23 of 38 April 21, 2017 Table 12: Summary of Core Neutron Source Sensitivity Study Case Number Description 1 Peripheral source strength biased by a factor of 0.95 2 Peripheral source strength biased by a factor of 1.05 3 Pin power distribution gradient diminished according to: =[( 1.0)x0.9]+1.0 4 Pin power distribution gradient intensified according to:
=[( 1.0)x1.1]+1.0 5 1 Mid-cycle burnup at 3000 MWD/MTU 6 1 Mid-cycle burnup at 50,000 MWD/MTU 7 Axial power distribution gradient intensified according to:
=[( 1.0)x1.1]+1.0 8 Axial power distribution gradient diminished according to:
=[( 1.0)x0.9]+1.0  Table 13: Source Permutation
-to-Nominal Fast Neutron (E > 1.0 MeV) Fluence Rate Difference at Surveillance Capsule Locations Case Number Surveillance Capsule Location 1 -4% 2 4% 3 1% 4 -1% 5 -7% 6 1% 7 1% 8 -1%
1 Cases 5 and 6 span a mid
-cycle burnup range of 47000 MWD/MTU. The uncertainty in the neutron fluence attributable to a 5000 MWD/MTU uncertainty in burnup is obtained by scaling the difference between Cases 5 and 6 accordingly by F = (5000 / 47000).
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 24 of 38 April 21, 2017 Table 14: Source Permutation
-to-Nominal Fast Neutron (E > 1.0 MeV) Fluence Rate Difference at Pressure Vessel Locations Case Number RPV Inside Radius
+42 cm Relative to Top-of-Core Elevation 2 RPV Inside Radius
+12 cm Relative to Middle-of-Core Elevation 2 RPV Inside Radius
-26 cm Relative to Bottom-of-Core Elevation 2 1 -4% -5% -4% 2 4% 5% 4% 3 0% 1% 0% 4 0% -1% 0% 5 -7% -7% -7% 6 4% 2% 3% 7 -10% 1% -8% 8 10% -1% 8%  Table 15: Source Permutation
-to-Nominal Fast Neutron (E > 1.0 MeV) Fluence Rate Difference at Reactor Cavity Locations Case Number Reactor Cavity  Top-of-Core Elevation Reactor Cavity Middle-of-Core Elevation Reactor Cavity  Bottom-of-Core Elevation 1 -4% -5% -5% 2 4% 5% 5% 3 1% 1% 1% 4 -1% -1% -1% 5 -7% -7% -7% 6 2% 2% 2% 7 -3% 1% -2% 8 3% -1% 2%
2 The selected locations on the inner radius of the reactor pressure vessel are typical of circumferential welds that join base metal forgings.
 
Westinghouse Non-Proprietary Class 3  LTR-REA-1 6-117, Rev. 2 Attachment 1, Page 25 of 38 April 21, 2017 Table 16: Summary of Neutron Fluence Rate Uncertainties at Surveillance Capsule Locations Resulting from Core Neutron Source Uncertainties Uncertainty Component Surveillance Capsule Location Peripheral Assembly Source Strength 4% Pin Power Distribution 1% Peripheral Assembly Burnup (+/-5000 MWD/MTU) 1% Axial Power Distribution 1%  Table 17: Summary of Neutron Fluence Rate Uncertainties at Pressure Vessel Locations Resulting from Core Neutron Source Uncertainties Uncertainty Component RPV Inside Radius
+42 cm Relative to Top-of-Core Elevation RPV Inside Radius
+12 cm Relative to Middle-of-Core Elevation RPV Inside Radius
-26 cm Relative to Bottom-of-Core Elevation Peripheral Assembly Source Strength 4% 5% 4% Pin Power Distribution 0% 1% 0% Peripheral Assembly Burnup (+/-5000 MWD/MTU) 1% 1% 1% Axial Power Distribution 10% 1% 8%  Table 18: Summary of Neutron Fluence Rate Uncertainties at Reactor Cavity Locations  Resulting from Core Neutron Source Uncertainties Uncertainty Component Reactor Cavity  Top-of-Core Elevation Reactor Cavity Middle-of-Core Elevation Reactor Cavity  Bottom-of-Core Elevation Peripheral Assembly Source Strength 4% 5% 5% Pin Power Distribution 1% 1% 1% Peripheral Assembly Burnup
(+/-5000 MWD/MTU) 1% 1% 1% Axial Power Distribution 3% 1% 2%
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 26 of 38 April 21, 2017 Analytic Uncertainty Analysis
- Geometric and Temperature Uncertainties  To assess the impact of uncertainties in the location and thickness of reactor components, as well as uncertainties in reactor coolant temperature, on calculated neutron fluence results, changes in the following parameters were evaluated:
Reactor internals dimensions - Thickness tolerances on stainless steel reactor internals components (e.g., core baffle, core barrel, thermal shield/neutron pad) are typically specified as 1/16 inch or tighter.
Reactor vessel inner radius - Reactor vessels typically specify an inner radius with tolerance bounds of -0.00 inches and +1/32 inches. A tolerance of +/- 1/8 inch is considered. Reactor vessel thickness - Some techniques for fabricating reactor vessels result in larger-than-nominal reactor vessel base metal plate thicknesses. A tolerance of +/- 1/16 inch is considered.
Dosimetry Positioning - Surveillance capsules have a tolerance of +/- 1/16 inch associated with the positioning of the dosimetry in radial, azimuthal, and axial directions. A larger positioning uncertainty of +/- 2 inches is associated with ex
-vessel neutron dosimetry in radial azimuthal, and axial directions. Coolant Temperature - Variations in water temperature over the course of a fuel cycle are expected to be less than +/- 10 °F.
Core Peripheral Modeling
- The modeling of the rectilinear core baffle in cylindrical geometry represents another potential source of uncertainty in the geometric modeling of the reactor. The sensitivity of the solution to the modeling approach is determined by a direct comparison of the results of a cylindrical geometry calculation with those of a Cartesian geometry calculation in which the baffle region and core periphery were modeled explicitly. The comparisons of interest were taken at various locations external to the core baffle, but inside the core barrel.
Each case evaluated as part of the sensitivity study is described in Table 19. The base case consisted of a low-leakage power distribution from a Westinghouse 4-Loop reactor. Table 20 through Table 22 provide the differences between calculated fast neutron (E > 1.0 MeV) fluence rate results at several locations for each permutation case, each normalized to the corresponding base case result. The overall uncertainty estimates are summarized in Table 23 through Table 25.
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 27 of 38 April 21, 2017 Table 19: Summary of Geometry and Temperature Sensitivity Study Case Number Description 1 Baffle plates, core barrel, and neutron pad thickness decreased by 1/16 inch 2 Baffle plates, core barrel, and neutron pad thickness increased by 1/16 inch 3 Reactor coolant temperatures decreased by 10 °F 4 Reactor coolant temperatures increased by 10
°F 5 Reactor vessel radius decreased by 1/8 inch 6 Reactor vessel radius increased by 1/8 inch 7 Reactor vessel thickness decreased by 1/16 inch 8 Reactor vessel thickness increased by 1/16 inch 9 Surveillance capsule position adjusted by 1/16 inch, ex-vessel dosimetry position adjusted by 2 inches 10 Cartesian versus cylindrical geometry modeling difference in core periphery Table 20: Geometry and Temperature Permutation
-to-Nominal Fast Neutron (E > 1.0 MeV) Fluence Rate Difference at Surveillance Capsule Locations Case Number Surveillance Capsule Location 1 1% 2 -1% 3 -4% 4 5% 5 0% 6 0% 7 0% 8 0% 9 2%(a) 10 5%(b) (a) Surveillance capsule positioning uncertainty includes radial, azimuthal, and axial position variations (b) Core periphery modeling uncertainty determined from direct comparison between cylindrical and Cartesian results in bypass region
 
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 28 of 38 April 21, 2017 Table 21: Geometry and Temperature Permutation
-to-Nominal Fast Neutron (E > 1.0 MeV) Fluence Rate Difference at Pressure Vessel Locations Case Number RPV Inside Radius
+42 cm Relative to Top-of-Core Elevation RPV Inside Radius
+12 cm Relative to Middle-of-Core Elevation RPV Inside Radius
-26 cm Relative to Bottom-of-Core Elevation 1 6% 5% 5% 2 -3% -3% -3% 3 -8% -6% -7% 4 10% 6% 8% 5 3% 4% 4% 6 -3% -4% -3% 7 2% -1% 1% 8 2% -1% 1% 9 N/A N/A N/A 10 5%* 5%* 5%*
* Core periphery modeling uncertainty determined from direct comparison between cylindrical and Cartesian results in bypass region   
 
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 29 of 38 April 21, 2017 Table 22: Geometry and Temperature Permutation
-to-Nominal Fast Neutron (E > 1.0 MeV) Fluence Rate Difference at Reactor Cavity Locations Case Number Reactor Cavity  Top-of-Core Elevation Reactor Cavity Middle-of-Core Elevation Reactor Cavity  Bottom-of-Core Elevation 1 3% 3% 3% 2 -4% -3% -3% 3 -6% -6% -6% 4 7% 6% 6% 5 1% 1% 1% 6 -4% -4% -4% 7 2% 2% 2% 8 -3% -3% -3% 9 12%(a) 4%(a) 14%(a) 10 5%(b) 5%(b) 5%(b) (a) Cavity capsule positioning uncertainty includes radial, azimuthal, and axial position variations (b) Core periphery modeling uncertainty determined from direct comparison between cylindrical and Cartesian results in bypass region Table 23: Summary of Neutron Fluence Rate Uncertainties at Surveillance Capsule Locations Resulting from Geometry and Temperature Uncertainties Uncertainty Component Surveillance Capsule Location Internals Dimensions 1% Vessel Inner Radius (IR) 0% Vessel Thickness 0% Dosimetry Position 2% Coolant Temperature 5% Core Periphery Modeling 5%
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 30 of 38 April 21, 2017 Table 24: Summary of Neutron Fluence Rate Uncertainties at Pressure Vessel Locations Resulting from Geometry and Temperature Uncertainties Uncertainty Component RPV Inside Radius
+42 cm Relative to Top-of-Core Elevation RPV Inside Radius
+12 cm Relative to Middle-of-Core Elevation RPV Inside Radius
-26 cm Relative to Bottom-of-Core Elevation Internals Dimensions 6% 5% 5% Vessel IR 3% 4% 4% Vessel Thickness 2% 1% 1% Coolant Temperature 10% 6% 8% Core Periphery Modeling 5% 5% 5%  Table 25: Summary of Neutron Fluence Rate Uncertainties at Reactor Cavity Locations Resulting from Geometry and Temperature Uncertainties Uncertainty Component Reactor Cavity  Top-of-Core Elevation Reactor Cavity Middle-of-Core Elevation Reactor Cavity  Bottom-of-Core Elevation Internals Dimensions 4% 3% 3% Vessel IR 4% 4% 4% Vessel Thickness 3% 3% 3% Dosimetry Position 12% 4% 14% Coolant Temperature 7% 6% 6% Core Periphery Modeling 5% 5% 5%   
 
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 31 of 38 April 21, 2017 Analytic Uncertainty Analysis - Summary Table 26 through Table 28 summarize the analytic uncertainties determined from a reference Westinghouse 4-Loop reactor model with calculations performed with RAPTOR-M3G. The total analytic uncertainty is derived by combining the individual uncertainty components in quadrature using the "root-sum-of-the-squares" method. The analytic uncertainty analysis was performed with both the TW and DTW differencing schemes. In general, the analytic uncertainty values are consistent between the two differencing schemes; however, in cases where there are differences, the higher uncertainty value was selected. This analytic uncertainty analysis addresses Regulatory Position 1.4.1 of Regulatory Guide 1.190. Table 26: Summary of Neutron Fluence Rate Uncertainties at Surveillance Capsule Locations Uncertainty Component Surveillance Capsule Location Peripheral Assembly Source Strength 4% Pin Power Distribution 1% Peripheral Assembly Burnup
(+/-5000 MWD/MTU) 1% Axial Power Distribution 1% Internals Dimensions 1% Vessel IR 0% Vessel Thickness 0% Dosimetry Position 2% Coolant Temperature 5% Core Periphery Modeling 5% Total Analytic Uncertainty 9%   
 
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 32 of 38 April 21, 2017 Table 27: Summary of Neutron Fluence Rate Uncertainties at Pressure Vessel Inner Radius Locations Uncertainty Component RPV Inside Radius
+42 cm Relative to Top-of-Core Elevation RPV Inside Radius
+12 cm Relative to Middle-of-Core Elevation RPV Inside Radius
-26 cm Relative to Bottom-of-Core Elevation Peripheral Assembly Source Strength 4% 5% 4% Pin Power Distribution 0% 1% 0% Peripheral Assembly Burnup
(+/-5000 MWD/MTU) 1% 1% 1% Axial Power Distribution 10% 1% 8% Internals Dimensions 6% 5% 5% Vessel IR 3% 4% 4% Vessel Thickness 2% 1% 1% Dosimetry Position N/A N/A N/A Coolant Temperature 10% 6% 8% Core Periphery Modeling 5% 5% 5% Total Analytic Uncertainty 17% 11% 15%
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 33 of 38 April 21, 2017 Table 28: Summary of Neutron Fluence Rate Uncertainties at Reactor Cavity Locations Uncertainty Component Reactor Cavity  Top-of-Core Elevation Reactor Cavity Middle-of-Core Elevation Reactor Cavity  Bottom-of-Core Elevation Peripheral Assembly Source Strength 4% 5% 5% Pin Power Distribution 1% 1% 1% Peripheral Assembly Burnup
(+/-5000 MWD/MTU) 1% 1% 1% Axial Power Distribution 3% 1% 2% Internals Dimensions 4% 3% 3% Vessel IR 4% 4% 4% Vessel Thickness 3% 3% 3% Dosimetry Position 12% 4% 14% Coolant Temperature 7% 6% 6% Core Periphery Modeling 5% 5% 5% Total Analytic Uncertainty 17% 12% 18%
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 34 of 38 April 21, 2017 Estimate of Bias and Uncertainty The simulator benchmark comparison results demonstrate that, when the configuration of the system is well-known, the level of agreement between RAPTOR
-M3G calculations and measurements is within the uncertainties associated with the measurements, themselves. Therefore no systematic bias is assigned to the calculational methodology. The following summarizes the uncertainties applicable to pressure vessel core-adjacent beltline locations, determined from the results of the methodology qualification process: Simulator Benchmark Comparisons 3% H. B. Robinson Benchmark Comparisons 8% Analytic Sensitivity Studies 11%  Peripheral Assembly Source Strength 5% Pin Power Distribution 1% Peripheral Assembly Burnup 1% Axial Power Distribution 1% Internals Dimensions 5% Vessel IR 4% Vessel Thickness 1% Coolant Temperature 6% Core Periphery Modeling 5% Other Factors 5%  Net Uncertainty 15%  The category designated "Other Factors" is intended to attribute an additional uncertainty to other geometrical or operational variables that individually have an insignificant effect on the overall uncertainty, but collectively should be accounted for in the assessment.
The uncertainty components tabulated above represent percent uncertainty at the level. In the tabulation, the net uncertainty from the analytic sensitivity studies has been broken down into its individual components. When the four uncertainty values listed above (3%, 8%, 11%, and 5%) are combined in quadrature, the resultant overall calculational uncertainty is estimated to be bounded by 15% for pressure vessel inner radius within the core-adjacent beltline region. This uncertainty quantification addresses Regulatory Position 1.4.3 of Regulatory Guide 1.190.
 
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 35 of 38 April 21, 2017 Operating Power Reactor Comparisons In addition to the uncertainty qualification comparisons described above, the radiation transport methodology in RAPTOR-M3G has been extensively compared with data from operating power reactors. These comparisons are intended to provide support for the validation of the transport calculation itself a s well as validation for the uncertainties assigned to the results of those calculations.
There are 69 in-vessel surveillance capsules with 295 threshold foil measurements from 18 nuclear power plants that have been analyzed with RAPTOR-M3G. In addition to the in-vessel surveillance capsules, 87 ex-vessel neutron dosimetry (EVND) capsules with 454 threshold foil measurements from locations in the reactor cavity opposite the core midplane have been analyzed with RAPTOR
-M3G. The average M/C reaction rate ratio over all the fast neutron sensors from each reactor is listed in Table 29. This tabulation provides a direct comparison, on an absolute basis, of measurement and calculation. For surveillance capsules, these comparisons show an average M/C ratio of 1.03 with a standard deviation of 5% at the level. For ex
-vessel dosimeters irradiated opposite the core midplane, these comparisons show an average M/C ratio of 0.92 with a standard deviation of 6% at the level. These results show that the M/C reaction rate ratios for the in
-vessel measurements are essentially unbiased and well within the +/- 20% acceptance criterion given in Regulatory Guide 1.190.
The M/C reaction rate ratios for the ex
-vessel measurements are within the +/- 30% criterion given in Regulatory Guide 1.190 for the cavity capsules.
This operating power reactor comparisons described in this section contribute to addressing Regulatory Position 1.4.2 of Regulatory Guide 1.190.
 
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 36 of 38 April 21, 2017 Table 29: In-Vessel and Ex
-Vessel Capsules Threshold Reactions M/C Reaction Rate Ratios Plant Number In-Vessel M/C EVND Midplane M/C Domestic Plant #1 1.05 0.96 Domestic Plant #2 0.99 0.97 International Plant #1 1.13 1.03 International Plant #2 1.06 1.00 International Plant #3 N/A 0.97 International Plant #4 0.99 0.89 International Plant #5 1.09 0.88 International Plant #6 0.95 0.87 International Plant #7 0.95 0.86 Domestic Plant #3 1.02 0.89 Domestic Plant #4 1.01 0.89 Domestic Plant #5 1.00 0.93 International Plant #8 0.96 0.87 International Plant #9 1.08 0.83 Domestic Plant #6 1.01 0.90 Domestic Plant #7 1.07 N/A Domestic Plant #8 1.11 N/A Domestic Plant #9 1.07 N/A Average 1.03 0.92 Std. Dev. %
5% 6% Total Number of Capsules 69 87 Total Number of Threshold Foils 295 454 Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 37 of 38 April 21, 2017 Measurement Data from Waterford Unit 3 The latest analysis of Waterford Unit 3 surveillance capsule dosimetry is described in detail in Reference 17 3. This analysis was performed with the RAPTOR
-M3G code. Anisotropic scattering was treated with a P 5 Legendre expansion, and angular discretization was modeled with an S 16 order of angular quadrature. The reactor vessel neutron exposure levels calculated as part of this analysis were used as inputs to the Waterford Unit 3 LRA submittal.
The comparison of the calculated results with the available plant
-specific dosimetry results was used solely to demonstrate the adequacy of the radiation transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was no t used to bias the final calculated neutron fluence results in any way.
Results of the evaluations of the dosimetry from the Waterford Unit 3 surveillance capsules withdrawn to date are provided in Table 30. Calculations of individual threshold sensor reaction rates are compared directly with the corresponding measurements. These threshold reaction rate comparisons provide a good evaluation of the accuracy of the fast neutron portion of the calculated energy spectra.
For the individual threshold foils, the average M/C comparisons for fast neutron reactions range from 1.07 to 1.17. The overall average M/C ratio for the entire set of Waterford Unit 3 data is 1.11 with an associated standard deviation of 7.0%.
These data comparisons show that the measurements and calculations agree within the 20% criterion specified in Regulatory Guide 1.190. Table 30: Comparison of M/C Sensor Reaction Rate Ratios  for Fast Neutron Threshold Reactions from Waterford Unit 3 Capsule M/C 63 46Ti(n,p) 54Fe(n,p) 58Ni(n,p) 238U(n,f) 97° 1.25 -4 1.07 1.06 -4 263° 0.98 1.15 1.05 1.11 -4 83° 1.05 1.19 1.10 1.17 -4 Average 1.09 1.17 1.07 1.11 - % Standard Deviation 12.8 2.4 2.3 4.9 - Average 1.11 % Standard Deviation 7.0 3 For brevity, this document does not include a discussion of the least squares adjustment process that is described in Reference 17, as it does not strictly pertain to the justification for the use of RAPTOR
-M3G. Note, however, that the least squares process in Reference 17 was used only as a check, and was not used to adjust the neutron exposure levels calculated for the LRA. 4 As part of the dosimetry analysis process, reaction rates for each foil were normalized to the measured Fe
-54 (n,p) reaction rate from each capsule. When from similar plants, they were discarded.
 
Westinghouse Non-Proprietary Class 3  LTR-REA-16-117, Rev. 2 Attachment 1, Page 38 of 38 April 21, 2017 References
: 1. Gulf General Atomic Report GA-8747, "TWOTRAN, a FORTAN Program for Two Dimensional Transport," July 1968.
: 2. WANL-PR-(LL)-034, "Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation. Vol. 5 - Two-Dimensional Discrete Ordinates Transport Technique," August 1970. 3. RSICC Computer Code Collection CCC-650, "DOORS 3.2a, One, Two-, and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System," Radiation Safety Information Computational Center, Oak Ridge National Laboratory (ORNL), May 2007.
: 4. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.
: 5. RSICC Data Library Collection DLC-185, "BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," Radiation Safety Information Computational Center, Oak Ridge National Laboratory (ORNL), July 1999.
: 6. RSICC Data Library Collection DLC-245, "VITAMIN-B7/BUGLE-B7, Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data," Radiation Safety Information Computational Center, Oak Ridge National Laboratory (ORNL), October 2011.
: 7. Westinghouse Report WCAP-17993-NP, Rev. 0-B, "Justification for the Use of RAPTOR-M3G for the Catawba Unit 1 Measurement Uncertainty Recapture (MUR) Power Uprate Fluence Evaluations," April 2015. (Available as ADAMS Accession Number ML15117A012.)
: 8. ORNL Report ORNL/TM-13205, "Pool Critical Assembly Pressure Vessel Facility Benchmark," (NUREG/CR-6454), July 1997. 9. RSICC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross Section Compendium," Radiation Shielding Information Computational Center, Oak Ridge National Laboratory, July 1994.
: 10. Nuclear Energy Agency (Organization for Economic Co-Operation and Development), "Prediction of Neutron Embrittlement in the Reactor Pressure Vessel: VENUS-1 and VENUS-3 Benchmarks," 2000.
: 11. ASTM Designation E844, 2009 (2014), "Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance," ASTM International, West Conshohocken, PA, 2014, DOI: 10.1520/E0844-09R14E01, www.astm.org. 12. ORNL Report ORNL/TM-13204, "H. B. Robinson-2 Pressure Vessel Benchmark,"
(NUREG/CR-6453), February 1998.
: 13. BNL Report BNL
-NUREG-52395, "PWR and BWR Pressure Vessel Fluence Calculation Benchmark Problems and Solutions," (NUREG/CR-6115), September 2001.
: 14. Westinghouse Report WCAP-13348, Rev. 0, "Consumers Power Company Palisades Nuclear Plant Reactor Vessel Fluence Analysis," May 1992.
: 15. Westinghouse Report WCAP-13362, Rev. 0, "Westinghouse Fast Neutron Exposure Methodology for Pressure Vessel Fluence Determination and Dosimetry Evaluation," May 1992. 16. R. E. Maerker, "Application of LEPRICON Methodology to LWR Pressure Vessel Surveillance Dosimetry," Reactor Dosimetry, Proc. 6th ASTM-Euratom Symposium, Jackson Hole, Wyoming, May 31 - June 5, 1987, American Society for Testing and Materials (1989).
: 17. Westinghouse Report WCAP-17969-NP, Rev. 0, "Analysis of Capsule 83° from the Entergy Operations, Inc. Waterford Unit 3 Reactor Vessel Radiation Surveillance Program," April 2015.
 
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 3 of 44 April 25, 2017 BackgroundIn the assessment of the state of embrittlement of reactor pressure vessels, an accurate evaluation of theneutron exposure for significant reactor pressure vessel materials is required. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", identifiesprocedures acceptable to the NRC staff for determining reactor pressure vessel neutron exposure(fluence). This Request for Additional Information (RAI) response provides supplemental information relating to the methodology used to determine neutron fluence at Waterford Unit 3.Neutron fluence has traditionally been quantified with discrete ordinates radiation transport calculations.The generically-approved Westinghouse fluence methodology, described in Section 2.2 of WCAP-14040-NP-A, Rev. 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCSHeatup and Cooldown Limit Curves," May 2004, discusses the use of either the two-dimensional (2D)
DORT code or the three-dimensional (3D) TORT code for calculating the transport of radiation from thereactor core to the reactor vessel wall and beyond.Westinghouse developed RAPTOR-M3G in order to overcome limitations associated with the TORTcode. RAPTOR-M3G is a 3D, parallel-processing discrete ordinates radiation transport code that followsessentially the same calculational methodology as TORT. The parallel-processing feature of RAPTOR-M3G allows large, 3D radiation transport calculations to be divided across networks of workstations andsolved simultaneously. This allows RAPTOR-M3G to perform calculations that would be prohibitivelytime consuming or impossible with TORT.Entergy has received a supplemental RAI from the NRC requesting further clarification concerning theapplication of the RAPTOR-M3G code, and also concerning uncertainties for neutron fluence calculationsapplicable to reactor vessel materials located outside the traditional "beltline" region. This letter providesthe requested information.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 4 of 44 April 25, 2017 NRC RAI 4.2.1-1a Request 1Provide information regarding the RAPTOR-M3G-code method in accordance with RG 1.190, Table 1and Section 3 (i.e., fluence calculation methods, fluence measurement methods, reporting provisions and their associated items described in RG 1.190, Table 1).Westinghouse Response to NRC RAI 4.2.1-1a Request 1Information regarding the fluence determination methodology employed for the Waterford Unit 3 LRA isavailable in LTR-REA-16-117, Rev. 2 (Reference 1). LTR-REA-16-117 addresses fluence calculationmethods and calculational methodology qualification. Measurement methods associated with the fluencedetermination methodology are described in this response. Documentation that meets the reporting provisions can be found in WCAP-17969-NP (Attachment 1 of Reference 2) and WCAP-18002-NP(Reference 3).The compliance matrix included at the end of this response identifies how the Waterford Unit 3 fluenceanalysis work complies with the provisions in Table 1 of Regulatory Guide 1.190 (Reference 4).
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 5 of 44 April 25, 2017Least Squares Adjustment with FERRETBackgroundLeast squares adjustment methods provide the capability to combine measurement data with the results ofneutron transport calculations to establish a best estimate neutron energy spectrum with associateduncertainties at the measurement locations. Best estimates for key exposure parameters such as neutronfluence rate,(E > 1.0 MeV), or iron atom displacement rate, dpa/s, along with their uncertainties arethen easily obtained from the adjusted spectrum. Using measurements in combination with detailedtransport calculations results in a reduced uncertainty in the calculated spectrum, and provides a methodto identify any biases or inconsistencies that may exist in the baseline transport calculation or in the measured data.The application of least squares adjustment methods in LWR dosimetry analysis is common throughoutthe dosimetry community. The ASTM International has addressed the use of adjustment codes in ASTME944 "Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance" (Reference 5), andmany industry workshops have been held to discuss the various applications. For example, the ASTM-EURATOM Symposia on Reactor Dosimetry periodically holds workshops on neutron spectrumunfolding and adjustment techniques at its conferences.In ASTM E944, a comprehensive listing of available adjustment codes commonly employed in reactorsurveillance programs including STAY'SL (Reference 6), LSL-M2 (Reference 7), LEPRICON(Reference 8), and FERRET is provided. Each of these codes is publicly available from the Radiation Safety Information Computational Center (RSICC) at ORNL.The FERRET code was initially developed at the Hanford Engineering Development Laboratory (HEDL)and has had extensive use in both the Liquid Metal Fast Breeder Reactor (LMFBR) program and theNRC-sponsored Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program(LWR-PV-SDIP). Examples of prior use of the FERRET code for LWR applications include: 1) a re-evaluation of the dosimetry from commercial pressurized water reactor (PWR) surveillance capsules(Reference 9) and 2) an evaluation of the dosimetry included in the PCA blind test experiments(Reference 10). Both of these applications were completed in support of the LWR-PV-SDIP program.The former evaluation was carried out to establish an updated surveillance capsule dosimetry database foruse in the establishment of improved trend curves defining the radiation induced shift in reference nil ductility transition temperature and the decrease in upper shelf energy versus neutron fluence ordisplacements per atom (dpa). These updated correlations were later used in the development of the trendcurve data appearing in Regulatory Guide 1.99, Revision 2 (References 11, 12, 13, and 14). The latter evaluation was completed to provide estimates of key exposure parameters (Fluence E > 1.0 MeV anddpa) for use in performing blind tests of neutron transport calculations in the PCA facility. This allowedblind test comparisons of calculated exposure parameters directly with the least squares results. It also allowed comparisons with measured reaction rates obtained from the multiple foil sensor sets included inthe PCA irradiations.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 6 of 44 April 25, 2017After participating in several cooperative efforts associated with the LWR-PV-SDIP, Westinghouseadopted the FERRET approach in the mid-1980s as the preferred approach to evaluating LWR surveillance dosimetry. The least squares methodology was deemed superior to the previously employedspectrum averaged cross-section approach, which is dependent on the accuracy of the shape of thecalculated neutron spectrum at the measurement locations. Further, applying the least squares methodology allowed for a rigorous treatment of uncertainties associated with dosimetry evaluation.Application of the Methodology In general, the least squares methods, as applied to LWR dosimetry evaluations, reconcile the measuredsensor reaction rate data, dosimetry reaction cross-sections, and calculated neutron energy spectra within their respective uncertainties. For example,+/-=+/-+/-(1)relates a set of measured reaction rates,, to a single neutron spectrum,, through the multigroupdosimeter reaction cross-section,, each with an uncertainty. The primary objective of the leastsquares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.The application of the least squares methodology requires the following input:
1.The calculated neutron energy spectrum and associated uncertainties at the measurement location.
2.The measured reaction rates and associated uncertainty for each sensor contained in the multiplefoil set.3.The energy-dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple foil sensor set.For LWR dosimetry applications, the calculated neutron spectrum is obtained from the results of plant-specific neutron transport calculations that follow the guidelines specified in Regulatory Guide 1.190. Thesensor reaction rates are derived from the measured specific activities obtained using established ASTM procedures. The dosimetry reaction cross-sections and uncertainties are obtained from the SNLRMLdosimetry cross-section library. Each of these critical input parameters and associated uncertainties arediscussed in this section.Input Spectrum UncertaintyThe neutron spectrum input to the least squares adjustment procedure is obtained directly from the resultsof plant-specific transport calculations for each sensor location. The spectrum at each location is input inan absolute sense (rather than as simply a relative spectral shape). Therefore, within the constraints of the Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 7 of 44 April 25, 2017assigned uncertainties, the calculated data are treated equally with the measurements. The inputuncertainties associated with the calculated spectrum must be consistent with the benchmarking results.The uncertainty matrix for the calculated spectrum is constructed from the following relationship:
=+(2)where specifies an overall fractional normalization uncertainty, and the fractional uncertainties and specify additional random groupwise uncertainties that are correlated with a correlation matrix given by:=[1]+(3)where=()2 (4)The first term in the correlation matrix equation specifies purely random uncertainties, while the secondterm describes the short-range correlations over a group range ( specifies the strength of the latterterm). The value of is 1.0 when g = and 0.0 otherwise.The normalization uncertainty pertains primarily to the magnitude of the spectrum, whereas the groupwise uncertainties pertain to the shape of the spectrum relative to energy. A typical set of parameters definingthe input uncertainties for the calculated spectrum is as follows:Fluence Rate Normalization Uncertainty ()15%Fluence Rate Group Uncertainties (,)(E > 0.0055 MeV) 15%(0.68 eV < E < 0.0055 MeV) 29%(E < 0.68 eV) 52%Short-Range Correlation ()(E > 0.0055 MeV) 0.9(0.68 eV < E < 0.0055 MeV) 0.5(E < 0.68 eV)
 
===0.5 Westinghouse===
Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 8 of 44 April 25, 2017Fluence Rate Group Correlation Range ()(E > 0.0055 MeV) 6(0.68 eV < E < 0.0055 MeV) 3(E < 0.68 eV) 2These uncertainty assignments provide an input covariance matrix that is consistent with the calculationaluncertainties defined through the benchmarking process.Measurement UncertaintyMeasurements at operating power reactors are generally accomplished with comprehensive multiple foilsensor sets including radiometric monitors (RM). In general, sensor sets employed in Westinghousedosimetry programs include materials in which the following reactions can be measured:In-VesselEx-Vessel Cu-63 (n,Co-60 Cu-63 (n,Co-60(Cd-covered)Ti-46 (n,p) Sc-46Ti-46 (n,p) Sc-46(Cd-covered)Fe-54 (n,p) Mn-54Fe-54 (n,p) Mn-54(Cd-covered)Ni-58 (n,p) Co-58Ni-58 (n,p) Co-58(Cd-covered)U-238 (n,f) FP(Cd-covered)U-238 (n,f) FP(Cd-covered)Np-237 (n,f) FP(Cd-covered)
Nb-93-93m (Cd-covered)
Co-59 (n,Co-60Np-237 (n,f) FP(Cd-covered)
Co-59 (n,Co-60(Cd-covered)
Co-59 (n,Co-60 Co-59 (n,Co-60(Cd-covered)These sensor sets provide adequate spectrum coverage in the fast neutron energy range greater thanapproximately 0.5 MeV and also include bare and cadmium covered cobalt sensors to provide anassessment of the thermal neutron fluence rate at the measurement locations. These sensor sets are fullyconsistent with the guidance specified in Section 2.1.1 of Regulatory Guide 1.190. Similar sensor set designs are also utilized by other vendors of LWR dosimetry programs.Following irradiation, the specific activity of each of the radiometric sensors is determined using the latestversion of ASTM counting procedures for each reaction. In particular, the following standards areapplicable to the radiometric sensors typically used in LWR programs:
E523Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Copper E526Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation ofTitanium E263Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Iron E264Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Nickel E704Standard Test Method for Measuring Reaction Rates by Radioactivation of Uranium-238E1297 Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Niobium Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 9 of 44 April 25, 2017 E705Standard Test Method for Measuring Reaction Rates by Radioactivation of Neptunium-237 E481Standard Test Method for Measuring Neutron Fluence Rates by Radioactivation of Cobalt andSilverE1005 Standard Test Method for Application and Analysis of Radiometric Monitors for ReactorVessel Surveillance E181Standard Test Methods for Detector Calibration and Analysis of RadionuclidesFollowing sample preparation and weighing, the specific activity of each sensor is determined using a germanium gamma spectrometer. In the case of multiple foil sensor sets, these analyses are usuallycompleted by direct counting of each of the individual sensors, or, as is sometimes the case with U-238and Np-237 fission monitors from in-vessel irradiations, by direct counting preceded by dissolution and chemical separation of cesium from the sensor.For ex-vessel dosimetry irradiations, gradient chains or wires are often included with the multiple foilsensor sets. For these gradient measurements, individual sensors are obtained by cutting the chains into aseries of segments to provide data at appropriate intervals over the extent of the beltline region of thepressure vessel. The determination of sensor specific activities in these segments then proceeds in the same fashion as for individual foils from the multiple foil sensor sets. In general, data from the followingreactions are obtained from the gradient chain measurements:Fe-54 (n,p) Mn-54Ni-58 (n,p) Co-58Co-59 (n,Co-60These data can be used in conjunction with high purity foil measurements to provide mappings of theneutron environment external to the reactor pressure vessel.For the radiometric sensors used in LWR irradiations, reaction rates referenced to full-power operation aredetermined from the following equation:
=1 (5)where:= measured specific activity (dps/g)= sensor reaction rate averaged over the irradiation period and referenced to operation ata core power level of (rps/atom)= number of target element atoms per gram of sensor (atom/g)
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 10 of 44 April 25, 2017= weight fraction of the target isotope in the target material= total number of monthly intervals comprising the irradiation period= number of product atoms produced per reaction= average core power level during irradiation period j (MW)= maximum or reference core power level of the reactor (MW)= calculated ratio of(E > 1.0 MeV) during irradiation period j to the time weightedaverage (E > 1.0 MeV) over the entire irradiation period= decay constant of the product isotope (s
-1)= length of irradiation period j (s)= decay time following irradiation period j (s)In the above equation, the ratio
/ accounts for month by month variation of power level within agiven fuel cycle. The ratio is calculated for each fuel cycle using the neutron transport methodologyand accounts for the change in sensor reaction rates caused by variations in fluence rate due to changes incore power spatial distributions from fuel cycle to fuel cycle. For a single cycle irradiation, = 1.0.However, for multiple fuel cycle irradiations, the additional correction must be utilized. This additionalcorrection can be quite significant for sensor sets that have been irradiated for many fuel cycles in areactor that has transitioned from non-low leakage to low leakage fuel management.Prior to using the measured reaction rates for direct comparison with the results of transport calculationsor as input to the least squares adjustment procedure, additional corrections must be made to U-238measurements to account for the presence of U-235 impurities in the sensors and for the build-in ofplutonium isotopes over the course of the irradiation. These corrections are location and fluence dependent, are obtained from a combination of calculated data from the plant-specific discrete ordinatesanalysis and, when available, measurements are made with U-235 foils.In addition to the corrections for competing neutron induced reactions in the U-238 sensors, correctionsmust also be made to both the U-238 and Np-237 sensor reaction rates to account for gamma-ray inducedfission reactions that occur during the irradiation. These photo-fission corrections are also location dependent and are obtained from the transport calculational methodology.Typical corrections to the measured fission rates at in-vessel and ex-vessel sensor locations aresummarized as follows:
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 11 of 44 April 25, 2017Typical CorrectionIn-VesselEx-VesselU-235 Impurities 8-12%<1%Pu Build-In7-20%<1%U-238 (,f)4-6%3-5%Np-237 (,f)1-2%1%The corrections listed are typical values for a PWR plant with in-vessel capsules mounted on the outerradius of the thermal shield. These values cannot be used with a plant-specific sensor set. Rather, theappropriate corrections must be determined for each sensor set based on the actual plant-specificgeometry and irradiation history.Along with the reaction rates, the uncertainty associated with each of these measurements is also animportant input to the least squares adjustment procedure. The overall uncertainty in the measured reaction rates includes components due to the basic measurement process, the irradiation historycorrections, and the corrections for competing reactions. A high level of accuracy in the reaction ratedeterminations is assured by using laboratory procedures that conform to the ASTM National Consensus Standards listed earlier in this section. In all cases, the latest available versions of the applicable standardsare used in the dosimetry evaluations.From these standards, the achievable uncertainties in the measured specific activities of each of thesensors comprising typical LWR multiple foil sensor sets are as follows:
ReactionPrecision Bias Cu-63 (n,) Co-60 1%3%Ti-46 (n,p) Sc-46 1%3%Fe-54 (n,p) Mn-54 1%3%Ni-58 (n,p) Co-58 1%3%U-238 (n,f) FP 1%5%Nb-93-93m 1%2%Np-237 (n,f) FP 1%5%Co-59 (n,) Co-60 3%0%These uncertainties included the effects of counting statistics, sample weighing, detector calibration,source/detector geometry corrections, and product nuclide branching ratios.In determining reaction rates from the measured specific activities, the following additional uncertainties are incurred:
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 12 of 44 April 25, 2017 ReactionFissionYield ProductHalf-Life Competing Reactions Cu-63 (n,) Co-60 0.01%Ti-46 (n,p) Sc-46 0.05%Fe-54 (n,p) Mn-54 0.06%Ni-58 (n,p) Co-58 0.08%U-238 (n,f) FP 3%0.30%4%Nb-93-93m 0.74%Np-237 (n,f) FP 4%0.30%1%Co-59 (n,) Co-60 0.01%After combing all of these uncertainty components, the sensor reaction rates derived from the countingand data evaluation procedures typically result in the following net uncertainties associated with the sensor reaction rates that are input to the least squares evaluation:
ReactionReaction RateUncertainty (1
)Cu-63 (n,) Co-60 5%Ti-46 (n,p) Sc-46 5%Fe-54 (n,p) Mn-54 5%Ni-58 (n,p) Co-58 5%U-238 (n,f) FP 10%Nb-93-93m 5%Np-237 (n,f) FP 10%Co-59 (n,) Co-60 5%In addition to the adherence to ASTM National Consensus Standards in the evaluation of sensor reaction rates, the procedures used by Westinghouse have been periodically tested via round robin countingexercises included as part of the LWR-PV-SDIP as well as by evaluation of fluence counting standardsprovided by the National Institute of Science and Technology (NIST). A summary of the results of these counting validations is as follows:
1980Round robin counting of the foil sets irradiated at the Thermal Shield Back (TSB)and Pressure Vessel Face (PVF) positions of the PCA simulator.
1981Round robin counting of additional foil sets included in the first metallurgical simulated surveillance capsule, also irradiated in the PCA benchmark mockup.These two counting exercises involved direct comparisons with measurements obtained by HEDL. At the Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 13 of 44 April 25, 2017time of these irradiations, HEDL was a prime contractor providing measurement services for the PCAbenchmark and was cross calibrated with NIST and the MOL laboratory in Belgium.
1985Counting and evaluation of Ti-46 (n,p) Sc-46, Fe-54 (n,p) Mn-54, andNi-58 (n,p) Co-58 certified fluence standards supplied by NIST.Comparisons with fluence standards involve the determination not only of the reaction rate of each foil, but also of the spectrum averaged cross-section in the NIST U-235 irradiation facility. Thus, the comparisons with certified fluence standards test both the measurement process and the energy dependentreaction cross-sections used in the evaluation.
1992Counting of NIST foils irradiated in an ex-vessel dosimetry experiment at the Trojanpower reactor.This exercise involved duplicate counting of a subset of irradiated foils by both Westinghouse and NISTto assure adequate cross-calibration of the laboratories so that data could be confidently mixed in theoverall fluence evaluations performed by NIST and ORNL.
1998Round robin counting of U-238 and Np-237 certified fluence standards irradiated byNIST in the MDRF facility at the University of Michigan.As in the case of the 1985 radiometric sensor evaluations, the fluence standard involved the determination of the reaction rate of each sensor, but also of the spectrum averaged cross-sections in the MDRF facility.The results obtained from these counting comparisons are summarized as follows:[West]/[HEDL][West]/[NIST]Average19801981198519921998 Cu-63 (n,) Co-601.014 1.018 0.969 1.000Ti-46 (n,p) Sc-46 1.0351.0121.030 1.026Fe-54 (n,p) Mn-540.9921.0081.0111.056 1.017Ni-58 (n,p) Co-581.0080.9901.0281.029 1.014U-238 (n,f) FP1.0191.0141.0171.017Np-237 (n,f) FP1.0101.0171.0951.041 Co-59 (n,) Co-601.013 1.017 1.015These comparisons demonstrate that the procedures used by Westinghouse in the determination of sensorreaction rates have produced accurate and stable results over an extended period of time. The cross-comparisons with HEDL and NIST support the reaction rate uncertainties used by Westinghouse inperforming LWR fluence evaluations.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 14 of 44 April 25, 2017In addition to these periodic comparisons, laboratory calibrations with NIST-supplied sources are alsocarried out on a routine basis.Dosimetry Cross-Sections and UncertaintyThe third key set of input data for the least squares procedure includes the reaction cross-sections for eachof the sensors included in the multiple foil dosimetry packages. The reaction rate cross-sections used byWestinghouse are taken from the SNLRML library (Reference 15). This data library provides reaction cross-sections and associated uncertainties, including covariances, for 66 dosimetry sensors in commonuse. Both the cross-sections and the uncertainties are provided in a fine multi-group structure for use inleast squares adjustment applications.These cross-sections were compiled from cross-section evaluations including ENDF/B-VI and IRDF-90(Reference 16) and have been tested with respect to their accuracy and consistency for least squares analyses. Further, the library has been empirically tested for use in fission spectra determination as well asin the fluence and energy characterization of 14 MeV neutron sources. Detailed discussions of thecontents of the SNLRML library along with the evaluation process for each of the sensors is provided in Reference 15.For the sensors of interest to LWR dosimetry applications, the following uncertainties in the fissionspectrum averaged cross-sections are provided in the SNLRML documentation package:
ReactionUncertainty Cu-63 (n,Co-60 4.08-4.16%Ti-46 (n,p) Sc-46 4.51-4.87%Fe-54 (n,p) Mn-543.05-3.11%Ni-58 (n,p) Co-58 4.49-4.56%U-238 (n,f) FP 0.54-0.64%
Nb-93-93m 6.96-7.23%Np-237 (n,f) FP10.32-10.97%
Co-59 (n,Co-60 0.79-3.59%These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with typical sensor sets used in LWR irradiations.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 15 of 44 April 25, 2017Validation of the Least Squares Adjustment ProceduresOverviewThis section describes the methodology qualification evaluations performed for the least squaresadjustment process. These evaluations consist of the following stages:
1.Data Comparisons in the NIST U-235 Fission Field: Comparisons of published spectrum-averaged cross-sections calculated with the SNLRML library to measurements from NIST, andcorresponding comparisons of Westinghouse-calculated spectrum-averaged cross-sections to datapublished in ASTM standards.
2.FERRET Sensitivity Studies:  Sensitivity studies examining the relative importance ofcommonly used neutron sensors to outputs of the least squares adjustment process and the effects of input uncertainties on final calculated uncertainties.
3.Operating Power Reactor Comparisons:An extensive database of evaluations performed withthe methodology described in this report. These comparisons are intended to provide support forthe validation of the transport calculation itself as well as validation for the uncertainties assignedto the results of those calculations.Data Comparisons in the NIST U-235 Fission FieldThe sensor reaction cross-sections and associated uncertainties play a key role in the least squaresevaluation of dosimetry data sets. Therefore, it is important to assess both the overall accuracy of thesecross-sections and the impact on that accuracy of any data processing included in the evaluation
 
procedure.The least squares approach used by Westinghouse makes use of the SNLRML dosimetry cross-sectionlibrary. This comprehensive library meets the standards established in ASTM E1018, "Standard Guide forApplication of ASTM Evaluated Cross-Section Data File" (Reference 17) for use in LWR applications,and has been explicitly recommended in previous versions of the standard. The library is provided by RSICC in a 640 neutron group format spanning an energy range from thermal to 20.0 MeV. Prior to use inthe least squares adjustment, this fine group library is collapsed to a broad group structure consisting of 53groups using the calculated neutron spectrum at the measurement location as a weighting function. The data comparisons from the standard field irradiations were used to determine the level of accuracy of thebase cross-section library, as well as to demonstrate the adequacy of the collapsing procedure used togenerate the 53 group library.In ASTM E261, "Standard Practice for Determining Neutron Fluence, Fluence Rate, and Spectra byRadioactivation Techniques" (Reference 18), fission spectrum averaged cross-sections applicable to the U-235 thermal fission field and the Cf-252 spontaneous fission field are provided for a variety ofthreshold activation detectors that are used in power and research reactor irradiations. In this datacompilation, both calculated and measured spectrum averaged cross-sections are provided along with their evaluated uncertainties. The magnitude of errors in the processed dosimetry cross-section library can Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 16 of 44 April 25, 2017be judged by the observed disagreement between the calculated spectrum averaged cross-sections and thecorresponding measured values for the standard U-235 and Cf-252 fields.The data listed in Table 1 and Table 2 have been extracted from Table 3 of ASTM E261, and arerepresentative of the foil sets used in power reactor irradiations and in the PCA benchmark irradiations.
This subset of the ASTM E261 information includes the threshold reactions typically used in LWRsurveillance capsule and ex-vessel dosimetry irradiations.For the comparisons shown in Table 1 and Table 2, the authors of ASTM E261 based the calculatedspectrum averaged reaction cross-sections on data from the recommended SNLRML library. Because theWestinghouse methodology and the evaluations provided in ASTM E261 are both based on the same dosimetry cross-section library, the calculated spectrum averaged cross-sections produced by theSAND/FERRET processing procedure should closely match the calculated values cited in the standard.Significant differences between the two sets of calculated spectrum averaged cross-sections would indicate errors in the processing procedure. Results of comparison of the calculated spectrum averagedcross-sections from ASTM E261 with the corresponding cross-sections processed by Westinghouse arelisted in Table 3.The calculation to measurement comparisons given in Table 1 indicate that for the U-235 thermal fissionfield, the C/M ratios except for Ti-46 (n,p) Sc-46 fall within one standard deviation of the combined uncertainty in the calculation and measurement. For these reactions, the agreement between calculationand measurement is within 5%. In the case of the Ti-46 (n,p) Sc-46 reaction, the C/M ratio falls withintwo standard deviations of the combined uncertainty with the calculation falling within 11% of the measured value.For the Cf-252 spontaneous fission field, the comparisons provided in Table 2 show that, with theexception of the U-238 (n,f) FP, In-115 (n,n') In-115m, and Ti-46 (n,p) Sc-46 reactions, the C/Mcomparisons fall within one standard deviation of the combined uncertainty in the calculations andmeasurements. The C/M ratios for the U-238 (n,f) FP and In-115 (n,n') In-115m reactions fall within two standard deviations and the C/M ratio for the Ti-46 (n,p) Sc-46 reaction falls within three standarddeviations of the combined uncertainty. For reactions other than Ti-46 (n,p) Sc-46, the agreement betweencalculation and measurement is within 6%. The calculated spectrum averaged cross-section for the Ti-46 (n,p) Sc-46 reaction falls within 11% of the measured value.The comparisons provided in Table 1 and Table 2 demonstrate that the SNLRML dosimetry cross-sections as processed by the authors of ASTM E261 produce accurate representations of the spectrumaveraged cross-sections in the NIST standard fission fields. The Westinghouse least squares approach usesthis same base dosimetry cross-section library, but a somewhat different processing procedure.To compare the cross-section processing procedure used in the Westinghouse approach to expand thecalculated input spectrum, spectrum weight the dosimetry cross-sections, and re-collapse the spectrum and dosimetry cross-sections to the FERRET 53 energy group structure, the ASTM E261 calculations forthe U-235 thermal fission field were duplicated for the foil reactions contained in both the power reactorsensor set and PCA sensor set. In performing this calculation, the ENDF/B-VI U-235 fission spectrum Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 17 of 44 April 25, 2017supplied with the BUGLE-96 cross-section library was input to the SAND/FERRET procedure as thecalculated spectrum. The dosimetry cross-sections were taken directly from the SNLRML library.Results of comparison of the Westinghouse-processed spectrum averaged cross-sections with thecalculated values from ASTM E261 are listed in Table 3 for both the power reactor and PCA sensor sets.
An examination of the data given in Table 3 shows that the spectrum averaged cross-sections calculatedby Westinghouse using the SAND pre-processing module are essentially identical to the calculated valuesgiven in ASTM E261, with the largest difference being at the 1% level.The comparison results summarized in Table 3 combined with the C/M results listed in Table 1 andTable 2 demonstrate that using the SNLRML dosimetry cross-section library combined with the algorithms included in the SAND pre-processing module to produce a spectrum weighted broad grouplibrary, results in an appropriate cross-section representation for use in the FERRET least squaresadjustment algorithm.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 18 of 44 April 25, 2017Table 1: U-235 Fission Spectrum Averaged Cross-Sections from ASTM E261Typical Power Reactor Sensor Sets ReactionSpectrum Average Cross-Section [millibarns]
C/MCalculationMeasurement Cu-63 (n,) Co-600.521 (2.85%, 6.05%)0.50 (11.0%)1.042 (12.87%)Ti-46 (n,p) Sc-4610.43 (2.46%, 5.4%)11.6 (3.45%)
0.899 (6.86%)Fe-54 (n,p) Mn-5480.18 (2.17%, 4.69%)80.5 (2.86%)0.996 (5.91%)Ni-58 (n,p) Co-58105.69 (2.43%, 4.52%)108.5 (5.0%)0.974 (7.16%)U-238 (n,f) FP306.23 (0.53%, 4.21%)309.0 (2.6%)0.991 (4.98%)
Nb-93-93m139.97 (3.06%, 4.14%)146.2 (8.6%)0.957 (10.02%)Np-237 (n,f) FP1330.1 (9.33%, 4.31%)1344.0 (4.0%)0.990 (11.0%)PCA Sensor Sets ReactionSpectrum Average Cross-Section [millibarns]
C/MCalculationMeasurement Al-27 (n,) Na-240.727 (1.40%, 6.95%)0.706 (3.97%)1.030 (8.13%)Ni-58 (n,p) Co-58105.69 (2.43%, 4.52%)108.5 (5.0%)0.974 (7.16%)In-115 (n,n') In-115m186.35 (2.17%, 4.17%)190.3 (3.84%)0.979 (6.07%)Rh-103 (n,n') Rh-103m706.02 (3.1%, 4.14%)733.0 (5.2%)0.963 (7.33%)U-238 (n,f) FP306.23 (0.53%, 4.21%)309.0 (2.6%)0.991 (4.98%)Np-237 (n,f) FP1330.1 (9.33%, 4.31%)1344.0 (4.0%)0.990 (11.0%)Notes: 1.The tabulated data were taken from Table 3 of ASTM E261.
2.For the calculated values, the cross-section and spectrum components of the uncertainty,respectively, are shown in parentheses.
3.The measurement uncertainty is also shown in parentheses.
4.The uncertainty in the C/M ratio represents a sum in quadrature of the measurement andcalculational uncertainty.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 19 of 44 April 25, 2017Table 2: Cf-252 Fission Spectrum Averaged Cross-Sections from ASTM E261Typical Power Reactor Sensor Sets ReactionSpectrum Average Cross-Section [millibarns]
C/MCalculationMeasurement Cu-63 (n,) Co-600.678 (2.83%, 1.38%)0.689 (1.98%)0.984 (3.72%)Ti-46 (n,p) Sc-4612.56 (2.45%, 1.18%)14.09 (1.76%)0.891 (3.24%)Fe-54 (n,p) Mn-5488.12 (2.14%, 0.79%)86.92 (1.34%)1.014 (2.65%)Ni-58 (n,p) Co-58115.31 (2.40%, 0.73%)117.6 (1.3%)
0.981 (2.83%)U-238 (n,f) FP315.39 (0.53%, 0.4%)325.0 (1.63%)0.970 (1.76%)
Nb-93-93m142.65 (3.04%, 0.36%)149.0 (7.0%)0.957 (7.64%)Np-237 (n,f) FP1335.0 (9.2%, 0.23%)1361.0 (1.58%)
0.981 (9.43%)PCA Sensor Sets ReactionSpectrum Average Cross-Section [millibarns]
C/MCalculationMeasurement Al-27 (n,) Na-241.04 (1.36%, 1.61%)1.017 (1.47%)1.019 (2.57%)Ni-58 (n,p) Co-58115.31 (2.40%, 0.73%)117.6 (1.3%)
0.981 (2.83%)In-115 (n,n') In-115m189.8 (2.16%, 0.38%)197.6 (1.3%)0.961 (2.55%)Rh-103 (n,n') Rh-103m714.45 (3.08%, 0.27%)757.0 (4.0%)0.944 (5.06%)U-238 (n,f) FP315.39 (0.53%, 0.4%)325.0 (1.63%)0.970 (1.76%)Np-237 (n,f) FP1335.0 (9.2%, 0.23%)1361.0 (1.58%)
0.981 (9.43%)Notes: 1.The tabulated data were taken from Table 3 of ASTM E261.
2.For the calculated values, the cross-section and spectrum components of the uncertainty,respectively, are shown in parentheses.
3.The measurement uncertainty is also shown in parentheses.
4.The uncertainty in the C/M ratio represents a sum in quadrature of the measurement andcalculational uncertainty.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 20 of 44 April 25, 2017Table 3: Comparison of Calculated U-235 Fission Spectrum Averaged Cross-SectionsTypical Power Reactor Sensor Sets ReactionSpectrum Average Cross-Section(millibarns)
Ratio FERRET/E261ASTM E261 SAND/FERRET Cu-63 (n,) Co-60 0.521 0.523 1.003Ti-46 (n,p) Sc-46 10.43 10.28 0.985Fe-54 (n,p) Mn-54 80.18 80.28 1.001Ni-58 (n,p) Co-58 105.69 105.76 1.001U-238 (n,f) FP 306.23 305.76 0.998 Nb-93-93m139.97 139.89 0.999Np-237 (n,f) FP1330.1141329.917 1.000PCA Sensor Sets ReactionSpectrum Average Cross-Section(millibarns)
Ratio FERRET/E261ASTM E261 SAND/FERRET Al-27 (n,) Na-24 0.727 0.729 1.003Ni-58 (n,p) Co-58 105.69 105.76 1.001In-115 (n,n') In-115m 186.35 186.17 0.999Rh-103 (n,n') Rh-103m 706.02 705.88 1.000U-238 (n,f) FP 306.23 305.76 0.998Np-237 (n,f) FP1330.1141329.917 1.000 Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 21 of 44 April 25, 2017FERRET Sensitivity StudiesThe information discussed in this subsection provides an understanding of how the composition of themultiple foil sensor sets and the input values for the uncertainties in the calculated neutron spectrum andmeasured reaction rates affect the results of the least squares analysis in terms of both the magnitude and uncertainty of the adjusted spectrum.The threshold foils comprising typical LWR sensor sets respond to different portions of the neutronenergy spectrum. These multiple foil measurements are a set of partial measurements of the fluence raterather than a group of complete and independent determinations. Of particular interest for weighting andaveraging threshold detector measurements is the spectrum coverage of the individual foils, which recognizes that such measurements do not form an equivalent observation set; therefore, they are noteasily matched to the principle of maximum likelihood.This response is shown in Figure 1 and Figure 2 for the neutron spectra characteristic of an in-vesselsurveillance capsule location and an ex-vessel dosimetry location. The graphical representations shown inFigure 1 and Figure 2 provide response profiles for the Cu-63 (n,Fe-54 (n,p), U-238 (n,f),Np-237 (n,f), and Nb-93 (n,n) threshold reactions, as well as for the neutron fluence rate (E > 1.0 MeV).The Ti-46 (n,p) and Ni-58 (n,p) reactions exhibit behavior similar to the Cu-63 (n,Fe-54 (n,p)reactions, respectively.From Figure 1 and Figure 2, it is evident that the response of the higher threshold reactions exhibitssignificantly different behavior than does the neutron fluence rate (E > 1.0 MeV); the fission monitor and niobium response shows a better match to the spectral behavior of the neutron fluence rate. This behaviorsuggests that in order to validate a calculation of the neutron fluence rate (E > 1.0 MeV), a significantspectral weighting of measured reaction rates should be included in the comparisons. The least squares approach allows for this spectral weighting to be included in a rigorous manner. The data from Figure 1and Figure 2 also indicate that the makeup of the foil set could affect the final results of the dosimetrycomparisons.FERRET Sensitivity Studies - Composition of the Multiple Foil Sensor SetTo assess the effect of the makeup of the sensor set on the least squares adjustment final solution,parametric studies were performed for typical LWR dosimetry data sets. In each parametric study, thecalculated neutron spectrum and uncertainty was held constant along with the uncertainties associated with the measured reaction rates. The base case consisted of an evaluation, including the thresholdreactions, whereas the variations in the analysis were accomplished by deleting foil reactions individuallyand in combination.The 11 cases analyzed in the parametric study with fission monitors are summarized in Table 4. Theresults of the least squares evaluations for each of these 11 cases are shown in Table 5. The 7 cases analyzed in the parametric study with niobium are summarized in Table 6. The results of the least squaresevaluations for each of these 7 cases are shown in Table 7.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 22 of 44 April 25, 2017This data in Table 5 show that in terms of the magnitude of the adjusted solution the results for all casesare within the uncertainty of the base case (Case 1). However, the uncertainty in the adjusted fluence rate increases as the content of the foil set is reduced with the highest uncertainties occurring when only asingle foil is used in conjunction with the transport calculation.The data in Table 5 further show that for a minimum uncertainty solution the foil set should consist of atleast Fe, U, and Np foils (Case 4). The addition of Cu, Ti, and Ni does not enhance the capability of thefoil set. This is because the very high threshold of the Cu-63 (n,and Ti-46 (n,p) reactions places theirresponse well above the important energy range for the neutron fluence rate (E > 1.0 MeV) and the Ni-58response is so similar to Fe-54 (n,p) that the reaction is redundant. Further reductions in the foil set fromthe Fe, U-238, Np-237 package results in an increased uncertainty in the adjusted fluence rate.As with the fission monitor sensitivity study, the data in Table 7 show that the uncertainty in the adjustedfluence rate increases as the content of the foil set is reduced. Similar to the fission monitor sensitivity study, a minimum uncertainty solution for the foil set should consist of at least Fe and Nb-93 foils (Case4). Further reductions in the foil set results in an increased uncertainty in the adjusted fluence rate.FERRET Sensitivity Studies - Input UncertaintiesA second sensitivity study was completed to evaluate the effect of the input uncertainties in the measuredreaction rates and the calculated neutron fluence rate on the adjusted solution and the final uncertaintiesdetermined by the FERRET least squares procedure. In performing this sensitivity study, the uncertaintymatrix in Table 8 was evaluated for the reaction rates and calculated spectra. In this sensitivity study, the "High" category would tend to overstate achievable uncertainties, the "Medium" category wouldrepresent a routinely achievable case, and the "Low" category would be equivalent to values achievable inthe laboratory or benchmark environment.The results of the input uncertainty sensitivity study are summarized in Table 9. Considering the"Medium"-"Medium" case as the baseline, the magnitude of the adjusted fluence rate varies by less than 2% for all of the cases evaluated, indicating that the magnitude of the adjusted fluence rate is dependentprimarily on the magnitude of the inputs, rather than on the input uncertainties. However, the associateduncertainty varies in a predictable trend from 4% for the "Low"-"Low" case to 11% for the "High"-
 
"High".The indications from this sensitivity study are that the FERRET least squares algorithm is operating asanticipated.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 23 of 44 April 25, 2017Table 4: Dosimeter Foil Composition Sensitivity Study Case Descriptions (with Fission Monitors)
CaseFoils Included in the FERRET Analysis Cu Ti Fe NiU-238Np-2371XXXXXX2XXXXX3X XXX 4 XXX 5 X X 6 X X 7XX 8 X 9 X 10 X11XTable 5: Dosimeter Foil Composition Sensitivity Study Results (with Fission Monitors)
Case(E > 1.0 MeV)[n/cm 2-s]A/C% DiffFrom Case 1 Calculated Adjusted 14.53E+10 (15%) 4.73E+10 (6%)1.04-24.53E+10 (15%) 4.73E+10 (6%)1.040.0%34.53E+10 (15%) 4.69E+10 (6%)1.04-0.8%44.53E+10 (15%) 4.69E+10 (6%)1.04-0.8%54.53E+10 (15%) 4.66E+10 (7%)1.03-1.4%64.53E+10 (15%) 4.70E+10 (7%)1.04-0.6%74.53E+10 (15%) 4.69E+10 (8%)1.04-0.7%84.53E+10 (15%) 4.61E+10 (12%)1.02-2.6%94.53E+10 (15%) 4.71E+10 (9%)1.04-0.4%104.53E+10 (15%) 4.67E+10 (9%)1.03-1.3%114.53E+10 (15%) 4.70E+10 (12%)1.04-0.6%Note: Numbers in parentheses represent one standard deviation.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 24 of 44 April 25, 2017Table 6: Dosimeter Foil Composition Sensitivity Study Case Descriptions (with Niobium)
CaseFoils Included in the FERRET Analysis Cu Ti Fe Ni Nb-931XXXXX2XXXX3X X X 4 X X 5 X 6 X7XTable 7: Dosimeter Foil Composition Sensitivity Study Results (with Niobium)
Case(E > 1.0 MeV)[n/cm 2-s]A/C% DiffFrom Case 1 Calculated Adjusted 14.73E+08 (15%) 4.79E+08 (6%)1.01-24.73E+08 (15%) 4.79E+08 (6%)1.010.0%34.73E+08 (15%) 4.88E+08 (6%)1.031.9%44.73E+08 (15%) 4.87E+08 (6%)1.031.5%54.73E+08 (15%) 4.92E+08 (8%)1.042.6%64.73E+08 (15%) 4.75E+08 (10%)1.01-0.9%74.73E+08 (15%) 4.46E+08 (13%)0.94-7.4%Note: Numbers in parentheses represent one standard deviation.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 25 of 44 April 25, 2017Table 8: Least Squares Adjustment Input Uncertainty Sensitivity StudyReaction Rate UncertaintiesReaction TypeUncertainty Category High Medium LowNon-Fission 10%5%2.5%Fission 20%10%5.0%Neutron Spectrum UncertaintiesReaction TypeUncertainty Category High Medium LowNormalization 30%20%10%Spectrum Groups 1-53 30%20%10%Spectrum Groups 28-4850%
25%10%Spectrum Groups 49-53100%
50%25%Table 9: Input Uncertainty Sensitivity Study Results -
Adjusted(E > 1.0 MeV) and % Standard Deviation SpectrumUncertaintyReaction Rate Uncertainty High Medium Low High4.72E+10 (11%)4.67E+10 (7%)4.66E+10 (5%)
Medium4.76E+10 (9%)4.70E+10 (6%)4.67E+10 (5%)
Low4.79E+10 (6%)4.76E+10 (5%)4.72E+10 (4%)
 
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 28 of 44 April 25, 2017Operating Power Reactor ComparisonsIn addition to the sensitivity studies described above, results of the radiation transport methodology inRAPTOR-M3G and the least squares dosimetry evaluations in FERRET have been extensively comparedwith data from operating power reactors. These comparisons are intended to provide support for the validation of the transport calculation itself as well as validation for the uncertainties assigned to theresults of those calculations.There are 69 in-vessel surveillance capsules with 295 threshold foil measurements from 18 nuclear powerplants that have been analyzed with RAPTOR-M3G and FERRET. The 18-reactor data set spansWestinghouse 2-Loop, 3-Loop, 4-Loop, and CE designs. In addition to the in-vessel surveillance capsules, 87 ex-vessel neutron dosimetry (EVND) capsules with 454 threshold foil measurements from locations inthe reactor cavity opposite the core midplane have been analyzed with RAPTOR-M3G and FERRET.The comparisons between the plant-specific calculations and the database measurements are provided ontwo levels. In the first instance, the average M/C reaction rate ratio over all the fast neutron sensors fromeach reactor is listed. This tabulation provides a direct comparison, on an absolute basis, of measurement and calculation. The results of this comparison are listed in Table 10. For surveillance capsules, thesecomparisons show an average M/C ratio of 1.03 with a standard deviation of 5% at the 1level. For ex-vessel dosimeters irradiated opposite the core midplane, these comparisons show an average M/C ratio of0.92 with a standard deviation of 6% at the 1level.The second level of comparison is based on the least squares adjustment procedure, which provides a weighting of the individual sensor measurements based on spectral coverage and allows a comparison ofthe neutron fluence rate (E > 1.0 MeV) before and after adjustment. The neutron fluence rate and fluence(E > 1.0 MeV) are the primary parameters of interest in the overall pressure vessel exposure evaluations.
The results are presented in Table 11 for surveillance capsules and Table 12 for ex-vessel dosimetry.The overall database average best estimate-to-calculation (BE/C) ratio for surveillance capsules is 0.98with an associated standard deviation of 6% for fast neutron fluence rate, and 0.99 with an associatedstandard deviation of 5% for iron atom displacement rate (DPA/s). For ex-vessel dosimetry, the averageBE/C ratio is 0.92 with an associated standard deviation of 6% for fast neutron fluence rate, and 0.93 with an associated standard deviation of 7% for DPA/s.These results show that the BE/C and M/C ratios are essentially unbiased and well within the +/- 20%acceptance criteria for the in-vessel capsules and +/- 30% for the cavity capsules given inRegulatory Guide 1.190.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 29 of 44 April 25, 2017Table 10: In-Vessel and Ex-Vessel Capsules Threshold Reactions M/C Reaction Rate RatiosPlant NumberIn-Vessel M/C EVNDMidplane M/CDomestic Plant #1 1.05 0.96Domestic Plant #2 0.99 0.97International Plant #1 1.13 1.03International Plant #2 1.06 1.00International Plant #3 N/A 0.97International Plant #4 0.99 0.89International Plant #5 1.09 0.88International Plant #6 0.95 0.87International Plant #7 0.95 0.86Domestic Plant #3 1.02 0.89Domestic Plant #4 1.01 0.89Domestic Plant #5 1.00 0.93International Plant #8 0.96 0.87International Plant #9 1.08 0.83Domestic Plant #6 1.01 0.90Domestic Plant #7 1.07 N/ADomestic Plant #81.11 N/ADomestic Plant #9 1.07 N/AAverage 1.03 0.92Std. Dev. %
5%6%Total Number of Capsules 69 87Total Number of Threshold Foils 295 454 Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 30 of 44 April 25, 2017Table 11: In-Vessel Surveillance Capsules BE/C Reaction Rate RatiosPlant NumberBE/C Fluence(E>1.0 MeV) BE/C DPANumber ofThreshold FoilsNumber of CapsulesInternational Plant #11.041.05 8 3International Plant #20.990.99 8 3International Plant #3N/AN/A N/A N/ADomestic Plant #11.031.04 20 4International Plant #40.920.90 14 4Domestic Plant #20.960.97 20 4International Plant #51.051.06 19 4International Plant #60.900.93 13 3International Plant #70.890.92 12 3Domestic Plant #30.970.99 30 6Domestic Plant #40.960.96 29 6Domestic Plant #50.930.95 25 5International Plant #80.940.95 14 4International Plant #91.041.04 16 4Domestic Plant #60.980.99 19 5Domestic Plant #71.021.03 13 3Domestic Plant #81.081.08 11 3Domestic Plant #91.021.02 24 5Average0.980.99295 (Total)69 (Total)Std. Dev. %6%5%
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 31 of 44 April 25, 2017Table 12: EVND Core Midplane BE/C Reaction Rate RatiosPlant Number(# of EVND Sets)BE/C Fluence(E>1.0 MeV) BE/C DPANumber ofThreshold FoilsNumber of CapsulesDomestic Plant #1 (1)0.930.90 15 4Domestic Plant #2 (1)0.920.88 16 4International Plant #1 (2)1.031.03 40 8International Plant #2 (2)1.001.01 40 8International Plant #3 (1)0.991.02 28 4International Plant #4 (1)0.890.88 20 4International Plant #5 (1)0.920.97 15 3International Plant #6 (2)0.890.91 48 8International Plant #7 (2)0.850.85 32 8Domestic Plant #3 (2)0.890.90 48 8Domestic Plant #4 (2)0.890.92 48 8Domestic Plant #5 (1)0.960.98 24 4International Plant #8 (2)0.860.88 40 8International Plant #9 (1)0.840.84 20 4Domestic Plant #6 (1)0.930.94 20 4Average0.920.93454 (Total)87 (Total)Std. Dev. %6%7%
Westinghouse Non-Proprietary Class 3LTR-REA-17-56,Rev. 1 Page 32 of 44 April 25, 2017Compliance MatrixRAPTOR-M3G Fluence Determination Regulatory Compliance Cross-ReferenceRegulatory Guide 1.190Regulatory PositionDescription of Regulatory Guidance Degree ofComplianceMapping/Reference toSatisfying DocumentationComments 1.3Fluence Determination. Absolute fluencecalculations, rather than extrapolated fluence measurements, must be used for the fluence determination.
FullRef. 1Ref. 2, Section 6 Ref. 3, Section 2Fluence rate values are determined directlyusing RAPTOR-M3G at the location of interest.1.1.1Modeling Data. The calculation modeling(geometry, materials, etc.) should be based ondocumented and verified plant-specific data.
FullRef. 1Ref. 2, Section 6 Ref. 3, Section 2Plant-specific data are used to develop all models analyzed with RAPTOR-M3G.1.1.2Nuclear Data. The latest version of the EvaluatedNuclear Data File (ENDF/B) should be used for determining nuclear cross-sections. Cross-section sets based on earlier or equivalent nuclear-data sets that have been thoroughly benchmarked are also acceptable. When the recommended cross-section data change, the effect of these changes on the licensee-specific methodology must be evaluated and the fluence estimates updated when the effects are significant.
FullRef. 1The BUGLE-96 library, based on the ENDF/B-VI nuclear data file, is used for RAPTOR-M3G calculations. BUGLE-96 has been superseded by the BUGLE-B7 library, which is derived from the newer ENDF/B-VII nuclear data file.
However, comparisons have conclusively demonstrated that the differences in fluence values derived from both libraries are very minor.1.1.2Cross-Section Angular Representation.
Indiscrete ordinates transport calculations, a P 3angular decomposition of the scattering cross-sections (at a minimum) must be employed.
FullRef. 1Ref. 2, Section 6 Ref. 3, Section 2The RAPTOR-M3G calculations performed for Waterford Unit 3 use P 5 cross-section data.1.1.2Cross-Section Group Collapsing. The adequacy ofthe collapsed job library must be demonstrated by comparing calculations for a representative configuration performed with both the master library and the job library.
FullRef. 19Validation work performed by Oak Ridge National Laboratory. The BUGLE-96 library has been in widespread use for neutron fluence calculations by analysts around the word for many years.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56,Rev. 1 Page 33 of 44 April 25, 2017RAPTOR-M3G Fluence Determination Regulatory Compliance Cross-ReferenceRegulatory Guide 1.190Regulatory PositionDescription of Regulatory Guidance Degree ofComplianceMapping/Reference toSatisfying DocumentationComments 1.2Neutron Source. The core neutron source shouldaccount for local fuel isotopics and, where appropriate, the effects of moderator density. The neutron source normalization and energy dependence must account for the fuel exposure dependence of the fission spectra, the number ofneutrons produced per fission, and the energyreleased per fission.
FullRef. 1The energy distribution of the source is determined by selecting a fuel burnuprepresentative of conditions averaged over theirradiation period under consideration and an initial fuel assembly enrichment characteristic of the core designs used over the applicable period.
From this average burnup, a fission split by isotope including U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242 is derived; and, from that fission split, composite values of energy release per fission, neutron yield per fission, and fission spectrum are determined.
1.2End-of-Life Predictions. Predictions of the vessel end-of-life fluence should be made with a best-estimate or conservative generic power distribution.
If a best estimate is used, the power distribution must be updated if changes in core loadings, surveillance measurements, or other information indicate a significant change in projected fluence
 
values.FullRef. 2, Section 6 Ref. 3, Section 2The future projections are based on the current reactor power level of 3716 MWt and include a 5% positive bias applied to the power generated in the peripheral fuel assemblies.1.3.1Spatial Representation. Discrete ordinatesneutron transport calculations should incorporate a detailed radial- and azimuthal-spatial mesh of ~2 intervals per inch radially. The discrete ordinates calculations must employ (at a minimum) an S 8quadrature and (at least) 40 intervals per octant.
FullRef. 1,Ref. 2, Section 6 Ref. 3, Section 2The geometric mesh description of the r, ,zreactor models consisted of 160 radial by 121 azimuthal by 247 axial intervals. Mesh sizes were chosen to ensure that proper convergence of the inner iterations was achieved on a pointwise basis. Angular discretization was modeled with an S 16 order of angularquadrature.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56,Rev. 1 Page 34 of 44 April 25, 2017RAPTOR-M3G Fluence Determination Regulatory Compliance Cross-ReferenceRegulatory Guide 1.190Regulatory PositionDescription of Regulatory Guidance Degree ofComplianceMapping/Reference toSatisfying DocumentationComments1.3.1Multiple Transport Calculations. If thecalculation is performed using two or more "bootstrap" calculations, the adequacy of the overlap regions must be demonstrated.Not ApplicableNot ApplicableBootstrapping was not performed.1.3.2 Point Estimates. If the dimensions of the tallyregion or the definition of the average-flux region introduce a bias in the tally edit, the Monte Carlo prediction should be adjusted to eliminate thecalculational bias. The average-flux regionsurrounding the point location should not include material boundaries or be located near reflecting, periodic, or white boundaries.Not ApplicableNot ApplicableThis requirement applies only to methodologies that use Monte Carlo techniques for radiation transport. RAPTOR-M3G does not use Monte Carlo techniques.1.3.2 Statistical Tests. The Monte Carlo estimated meanand relative error should be tested and satisfy all
 
statistical criteria.Not ApplicableNot ApplicableThis requirement applies only to methodologies that use Monte Carlo techniques for radiation transport. RAPTOR-M3G does not use Monte Carlo techniques.1.3.2Variance Reduction. All variance reductionmethods should be qualified by comparison with calculations performed without variance reduction.Not ApplicableNot ApplicableThis requirement applies only to methodologies that use Monte Carlo techniques for radiation transport. RAPTOR-M3G does not use Monte Carlo techniques.1.3.3Capsule Modeling. The capsule fluence isextremely sensitive to the geometrical representation of the capsule geometry and internal water region, and the adequacy of the capsule representation and mesh must be demonstrated.
FullThis Document (Operating Power ReactorComparisons)Ref. 2, Section 6 Ref. 3, Section 2The quality and quantity of the comparisons of measurements to RAPTOR-M3G calculations demonstrates the adequacy of the modeling.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56,Rev. 1 Page 35 of 44 April 25, 2017RAPTOR-M3G Fluence Determination Regulatory Compliance Cross-ReferenceRegulatory Guide 1.190Regulatory PositionDescription of Regulatory Guidance Degree ofComplianceMapping/Reference toSatisfying DocumentationComments1.3.3Spectral Effects on RTNDT. In order to account forthe neutron spectrum dependence of RT NDT, when itis extrapolated from the inside surface of the pressure vessel to the T/4 and 3T/4 vessel locations using the E > 1-MeV fluence, a spectral lead factor must be applied to the fluence for the calculation of NDT.FullRef. 2, Section 1Ref. 3, Section 4 - Section 6Ref. 11, Position 1.1The Westinghouse fluence determination methodology uses Equation (3) in Ref. 11. Page 13 in RG 1.190 states that, "this spectral lead factor has been included in the Equation 3attenuation formula of Revision 2 of RegulatoryGuide 1.99, and therefore is not required when this formula is used."1.3.5Cavity Calculations. In discrete ordinatestransport calculations, the adequacy of the S 8angular quadrature used in cavity transportcalculations must be demonstrated.
FullRef. 2, Section 6 Ref. 3, Section 2An S 8 angular quadrature set (with 80 discretedirections) was not used. Angular discretization for Waterford Unit 3 was modeled with an S 16order of angular quadrature (with 288 discrete directions).1.4.1, 1.4.2, 1.4.3Methods Qualification. The calculationalmethodology must be qualified by both (1)comparisons to measurement and calculational benchmarks and (2) an analytic uncertainty analysis. The methods used to calculate the benchmarks must be consistent (to the extent possible) with the methods used to calculate the vessel fluence. The overall calculational bias and uncertainty must be determined by an appropriate combination of the analytic uncertainty analysis and the uncertainty analysis based on the comparisons to the benchmarks.
FullRef. 1Complete details are available in Ref. 1.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56,Rev. 1 Page 36 of 44 April 25, 2017RAPTOR-M3G Fluence Determination Regulatory Compliance Cross-ReferenceRegulatory Guide 1.190Regulatory PositionDescription of Regulatory Guidance Degree ofComplianceMapping/Reference toSatisfying DocumentationComments1, 1.4.3Fluence Calculational Uncertainty.
The vesselfluence (1 sigma) calculational uncertainty must be demonstrated to be20% for RT PTS and RT NDTdetermination. In these applications, if the benchmark comparisons indicate differences greater than 20%, the calculational model must be adjusted or a correction must be applied to reduce thedifference between the fluence prediction and theupper 1-sigma limit to within 20%. For other applications, the accuracy should be determined using the approach described in Regulatory Position 1.4, and an uncertainty allowance should be included in the fluence estimate as appropriate in the specific application.
FullRef. 1Complete details are available in Ref. 1.2.1.1Spectrum Coverage. The set of dosimeters shouldprovide adequate spectrum coverage.
FullThis Document (FERRET Sensitivity Studies)Ref. 2, Appendix AFigure 1 compares the cumulative sensor response as a function of energy for in-vessel measurement locations. Specific energy response ranges of the sensors employed at Waterford Unit 3 can be found in Ref. 2, Table A-1 and Table A-2.2.1.1Dosimeter Nuclear and Material Properties.
Useof dosimeter materials should address melting, oxidation, material purity, total and isotopic mass assay, perturbations by encapsulations and thermal shields, and accurate dosimeter positioning.
Dosimeter half-life and photon yield and interference should also be evaluated.
FullNot ApplicableThese are requirements that apply to the design of new measurement programs.2.1.2Corrections. Dosimeter-response measurementsshould account for fluence rate variations, isotopic burnup effects, detector perturbations, self-shielding, reaction interferences, and photofission.
FullThis Document (Measurement Uncertainty)Ref. 2, Section 6Specific corrections employed for Waterford Unit 3 are described in Ref. 2, Section 6.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56,Rev. 1 Page 37 of 44 April 25, 2017RAPTOR-M3G Fluence Determination Regulatory Compliance Cross-ReferenceRegulatory Guide 1.190Regulatory PositionDescription of Regulatory Guidance Degree ofComplianceMapping/Reference toSatisfying DocumentationComments2.1.3Response Uncertainty. An uncertainty analysismust be performed for the response of each dosimeter.
FullThis Document (FERRET Sensitivity Studies)This information is provided in this document.
2.2Validation. Detector-response calibrations must becarried out periodically in a standard neutron field.
FullThis Document (Data Comparisons in the NIST U-235Fission Field)This information is provided in this document.
2.3Fast-Neutron Fluence. The E > 1 MeV fast-neutron fluence for each measurement location must be determined using calculated spectrum-averagedcross-sections and individual detectormeasurements. As an alternative, the detector responses may be used to determine reaction probabilities or average reaction rates.
FullRef. 2, Appendix ADetector responses are used to determine average reaction rates. The least-squares adjustment process produces a best-estimate neutron fluence (E > 1.0 MeV) value at the measurement location.
2.3Measurement-to-Calculation Ratios. The M/Cratios, the standard deviation and bias between calculation and measurement, must be determined.
FullRef. 2, Appendix AThis information is available in Ref. 2, Appendix A.
3.1Neutron Fluence and Uncertainties. Details ofthe absolute fluence calculations, associated methods qualification and fluence adjustments (if any) should be reported. Justification and a description of any deviations from the provisions of this guide should be provided.
FullRef. 2, Section 6 Ref. 3, Section 2This information is available in Ref. 2 and Ref. 3.3.2Neutron Fluence and Uncertainties.
Calculatedmultigroup neutron fluences and fluence rates should be reported.
FullRef. 2, Section 6 Ref. 3, Section 2Multigroup data is available in supporting documentation.
3.2Neutron Fluence and Uncertainties. The valueand basis of any bias or model adjustment made to improve the measurement-to-calculation agreement must be reported.Not ApplicableNot ApplicableWestinghouse methodology does not apply any bias to the calculated fast neutron fluence. The measurement data is only used to validate that the calculated fast neutron fluence is within
+/-20% of the measured data.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56,Rev. 1 Page 38 of 44 April 25, 2017RAPTOR-M3G Fluence Determination Regulatory Compliance Cross-ReferenceRegulatory Guide 1.190Regulatory PositionDescription of Regulatory Guidance Degree ofComplianceMapping/Reference toSatisfying DocumentationComments 3.3Neutron Fluence and Uncertainties.
Calculatedintegral fluences and fluence rates for E > 1 MeV and their uncertainties should be reported.
FullRef. 2, Section 6 Ref. 3, Section 2This information is available in Ref. 2 and Ref. 3.3.4Neutron Fluence and Uncertainties. Measuredand calculated integral E > 1 MeV fluences or reaction rates and uncertainties for each measurement location should be reported. The M/Cratios and the spectrum averaged cross-sectionshould also be reported.
FullRef. 2, Appendix AThis information is available in Ref. 2,Appendix A.
3.5Neutron Fluence and Uncertainties. The resultsof the standard field validation of the measurementmethod should be reported.
FullThis Document (Data Comparisons in the NIST U-235Fission Field)This information is provided in this document.
3.5Specific Activities and Average Reaction Rates.The specific activities at the end of irradiation andmeasured average reaction rates with uncertainties should be reported.
FullRef. 2, Appendix AThis information is available in Ref. 2, Appendix A.
3.5Specific Activities and Average Reaction Rates.All corrections and adjustments to the measuredquantities and their justification should be reported.
FullRef. 2, Appendix AThis information is available in Ref. 2, Appendix A.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 39 of 44 April 25, 2017 NRC RAI 4.2.1-1a Request 1 References 1.Westinghouse Letter LTR-REA-16-117, Rev. 2, "Response to the NRC Request for AdditionalInformation Regarding RAPTOR-M3G on the Waterford Unit 3 License Renewal Application,"
 
April 21, 2017.
2.Entergy Operations, Inc. Letter W3F1-2015-0056, "Submittal of Reactor Vessel Material Surveillance Program Capsule Test Results, Waterford Steam Electric Station, Unit 3(Waterford 3), Docket No. 50-382, License No. NPF-38" August 6, 2015. (Available as ADAMSAccession Number ML15222A360. See also ADAMS Accession Number ML15222A373 for alisting of enclosures.)
3.Westinghouse Report WCAP-18002-NP, Rev. 0, "Waterford Unit 3 Time-Limited AgingAnalysis on Reactor Vessel Integrity," July 2015.
4.Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure VesselNeutron Fluence," U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research,March 2001.
5.ASTM Designation E944, 2013, "Standard Guide for Application of Neutron SpectrumAdjustment Methods in Reactor Surveillance," ASTM International, West Conshohocken, PA,2013, DOI: 10.1520/E0944-13, www.astm.org.
6.RSICC Code Package PSR-113, "STAY'SL Least Squares Dosimetry Unfolding Code System,"
Radiation Shielding Information Computational Center, Oak Ridge National Laboratory (ORNL),December 1991.
7.RSICC Code Package PSR-233, "LSL-M2 Least Squares Logarithmic Adjustment of NeutronSpectra," Radiation Shielding Information Computational Center, Oak Ridge National Laboratory(ORNL), February 1991.
8.RSICC Code Package PSR-277, "LEPRICON PWR Pressure Vessel Surveillance Dosimetry Analysis System," Radiation Shielding Information Computational Center, Oak Ridge National Laboratory (ORNL), June 1995.
9.R. L. Simons, et. al, "Re-Evaluation of the Dosimetry for Reactor Pressure Vessel Surveillance Capsules," in NUREG/CP-0029, "Proceedings of the Fourth ASTM-EURATOM Symposium onReactor Dosimetry," F. B. K. Kam, Editor, July 1982.
10.W. N. McElroy, et. al, "LWR PV Surveillance Dosimetry Improvement Program: PCAExperiments and Blind Test," NUREG/CR-1861, July 1981.
11.Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U. S.
Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, May 1988.
12.G. L. Guthrie, "Uncertainty Considerations in Development and Application of Charpy Trend Curve Formulas," in REACTOR DOSIMETRY Dosimetry Methods for Fuels, Cladding, and Structural Materials, Proceedings of the Fifth ASTM-Euratom Symposium on Reactor Dosimetry, GKSS Research Centre, Geesthact, F. R. G., J. P Genthon and H Rottger, Editors, D. Reidel Publishing Company, 1985.
13.G. L. Guthrie, "Charpy Trend Curve Development Based on PWR Surveillance Data," in NUREG/CP-0048 Proceedings of the U. S. Nuclear Regulatory Commission Eleventh WaterReactor Safety Research Information Meeting, Volume 4 - Materials Engineering Research,"U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, January 1984.
14.G. L. Guthrie, "Charpy Trend Curves Based on 177 PWR Data Points," NUREG/CR-3391,Volume 2, LWR-PV-SDIP Quarterly Progress Report, April 1983.
15.RSICC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross Section Compendium," Radiation Shielding Information Computational Center, Oak Ridge NationalLaboratory, July 1994.
16.N. P. Kocherov, et. al, "International Reactor Dosimetry File (IRDF-90)," International AtomicEnergy Agency, Nuclear Data Section, IAEA-NDS-141, Rev. 0, August 1990.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 40 of 44 April 25, 2017 17.ASTM Designation E1018, 2013, "Standard Guide for Application of ASTM Evaluated CrossSection Data File, Matrix E706 (IIB)," ASTM International, West Conshohocken, PA, 2013,DOI: 10.1520/E1018-09R13, www.astm.org.
18.ASTM Designation E261, 2015, "Standard Practice for Determining Neutron Fluence, Fluence Rate, and Spectra by Radioactivation Techniques," ASTM International, West Conshohocken, PA, 2015, DOI: 10.1520/E0261-15, www.astm.org.
19.RSIC Data Library Collection DLC-185, "BUGLE-96, Coupled 47 Neutron, 20 Gamma RayGroup Cross-Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure VesselDosimetry Applications," Radiation Shielding Information Center, Oak Ridge NationalLaboratory (ORNL), March 1996.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 41 of 44 April 25, 2017 NRC RAI 4.2.1-1a Request 3The following items can affect fluence calculations for reactor vessel nozzle areas. Therefore, explainhow the following items were considered as part of the fluence calculational methodology:
a.Axial distribution and isotopic content of uranium-235 and plutonium-239 in the fuel when calculating the fission source b.Biological shield concrete composition c.Cavity gap between the reactor vessel and the biological shield d.Homogenized materials above and below the active core region e.Discretization effects on deterministic calculations (e.g., nozzle flux increases have resulted from changing from level symmetric (S
: 16) to quadruple range (QR
: 16) quadrature)Westinghouse Response to NRC RAI 4.2.1-1a Request 3aThe analysis considers spatial and spectral variations in the neutron source, derived from detailed assembly burnup distributions from individual fuel cycles. The source is spatially shaped by radial pingradients for fuel assemblies located on the core periphery, and an axial power distribution representativeof mid-cycle operating conditions.The energy distribution of the source is determined by selecting a fuel burnup representative of conditionsaveraged over the irradiation period under consideration and an initial fuel assembly enrichmentcharacteristic of the core designs used. From the assembly burnup and initial U-235 enrichment, a fissionsplit by isotope including U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242 is derived; and, from thatfission split, composite values of energy release per fission, neutron yield per fission, and fission neutronenergy spectrum are determined.Both the spatial shape of the axial power distribution and the energy spectrum of the source were treatedin the analytic uncertainty analysis; see Reference 1. The following methods were employed to assess theeffects of uncertainties in these inputs:Burnup of the peripheral fuel assemblies - Perturbations in fuel assembly burnup impact thefission neutron energy spectrum, neutron yield per fission, and energy released per fission for each peripheral fuel assembly. A 5000 MWD/MTU uncertainty in the peripheral fuel assemblyburnups is considered conservative. The sensitivity study is performed using a series ofcalculations starting with mid-cycle burnup at 3000 MWD/MTU, and 5000 MWD/MTU to50,000 MWD/MTU with 5000 MWD/MTU delta mid-cycle burnup between each run.Axial power distribution - Based on variations in axial peaking factors over the course of a fuelcycle, a 10% uncertainty in the shape of the axial power distribution is considered conservative.
This estimate was derived from a review of numerous axial distributions from a wide variety ofpressurized water reactors employing both low leakage and non-low leakage fuel management.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 42 of 44 April 25, 2017The Waterford Unit 3 upper-middle girth weld (106-121) is located at the reactor vessel inner radius at anelevation 52 cm relative to the top of the active fuel. The analytic uncertainty analysis was performedgenerically, and considered a location 42 cm above the top of the active fuel. The following uncertaintycomponents were attributed as a result of the analytic uncertainty analysis:Uncertainty ComponentRPV Inside Radius+42 cm Relative toTop-of-Core ElevationPeripheral Assembly Burnup (+/-5000 MWD/MTU) 1%Axial Power Distribution 10%Westinghouse Response to NRC RAI 4.2.1-1a Request 3bIn the modeling of the reactor cavity region a standard, "generic" concrete composition was assumed.Biological shield concrete composition was not considered in the analytic uncertainty analysis describedin Reference 1 because it was deemed unlikely to significantly affect the calculated values of fast neutron(E > 1.0 MeV) fluence for materials near or above the 1.00E+17 n/cm 2 threshold. Note that the analyticuncertainty estimate includes an "other factors" uncertainty component (5%), intended to addressgeometrical or operational variables that individually have an insignificant effect on the overalluncertainty, but collectively should be accounted for in the assessment.A concrete hydrogen number density of 7.8E-03 atoms/b-cm is assumed in the concrete mixture. This isnear the low end of the hydrogen content range considered by Oak Ridge National Laboratory (ORNL) intheir recent study of extended beltline neutron fluence calculations (Reference 2). The ORNL resultssuggest that lower hydrogen content would tend to provide conservative results relative to assumedconcrete compositions with higher hydrogen content.Westinghouse Response to NRC RAI 4.2.1-1a Request 3cPlant drawings showing as-designed/as-built reactor cavity dimensions were used in the construction ofthe radiation transport models.Reactor cavity air gap dimension was not considered in the analytic uncertainty analysis described inReference 1. The effect of this air gap is accounted for in the radiation transport calculations, and anyuncertainty in this value was not deemed a significant contributor to the calculated values of fast neutron(E > 1.0 MeV) fluence for materials near or above the 1.00E+17 n/cm 2 threshold. Note that the analyticuncertainty estimate includes an "other factors" uncertainty component (5%), intended to addressgeometrical or operational variables that individually have an insignificant effect on the overalluncertainty, but collectively should be accounted for in the assessment.Westinghouse Response to NRC RAI 4.2.1-1a Request 3dThe radiation transport model used an approximate mixture of stainless steel and borated water forregions directly above and below the core.The effect of varying assumptions in the balance between water and steel was not considered in theanalytic uncertainty analysis described in Reference 1. The model that formed the basis of the analytic Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 43 of 44 April 25, 2017uncertainty analysis used material mixtures derived from drawings of the reactor internals, and anyuncertainties in these values were not deemed significant contributors to the calculated values of fastneutron (E > 1.0 MeV) fluence for materials near or above the 1.00E+17 n/cm 2 threshold. Note that theanalytic uncertainty estimate includes an "other factors" uncertainty component (5%), intended to address geometrical or operational variables that individually have an insignificant effect on the overall uncertainty, but collectively should be accounted for in the assessment.Westinghouse Response to NRC RAI 4.2.1-1a Request 3eThe radiation transport model used a cylindrical mesh grid with 160 radial, 121 azimuthal, and 247 axialintervals. Anisotropic scattering was treated with a P 5 Legendre expansion and the angular discretizationwas modeled with an S 16 order of angular quadrature.Appendix F of WCAP-18060-NP, Rev. 1 (Reference 4) compares calculations performed with S 8quadrature and S 12 quadrature sets for three locations on the inner radius of the reactor vessel. Reactorvessel Weld W06 is located 42 cm above the elevation of the top of the active core, Weld W05 is located12 cm above the middle of the core, and Weld W04 is located 26 cm below the bottom of the core.Appendix F of Reference 4 shows no significant difference between calculated values of fast neutron(E > 1.0 MeV) fluence rate for any of the materials evaluated. Because negligible differences wereobserved when moving from S 8 to S 12 quadrature sets, increasing the quadrature order to S 16 is deemedunlikely to affect the calculated results for the similar calculation performed for Waterford Unit 3.The analytic uncertainty analysis documented in Reference 1 did not specifically evaluate discretizationeffects because they were considered unlikely to significantly affect the calculated values of fast neutron(E > 1.0 MeV) fluence for materials near or above the 1.00E+17 n/cm 2 threshold. Note that the analyticuncertainty estimate includes an "other factors" uncertainty component (5%), intended to address geometrical or operational variables that individually have an insignificant effect on the overall uncertainty, but collectively should be accounted for in the assessment.The judgment that discretization effects are unlikely to significantly affect calculated neutron fluenceresults is supported by the results in Reference 2. The reactor vessel materials likely to be near or abovefast neutron (E > 1.0 MeV) fluence levels of 1.00E+17 n/cm 2 in the extended beltline region are thebottom extents of the inlet and outlet nozzle welds. The Reference 2 analysis (slide 53) indicates that thedifference between using QR 16 and S 16 quadrature in these regions is negligible; therefore, a 5%uncertainty for "other factors" sufficiently encompasses any effect that could be imparted by different quadrature settings.
NRC RAI 4.2.1-1a Request 3 References 1.Westinghouse Letter LTR-REA-16-117, Rev. 2, "Response to the NRC Request for AdditionalInformation Regarding RAPTOR-M3G on the Waterford Unit 3 License Renewal Application,"
April 21, 2017.
2.NRC Letter, "Summary of U.S. Nuclear Regulatory Commission Computation of NeutronFluence Information Exchange Public Meeting," February 7, 2017. (Available as ADAMSAccession Number ML17038A134. The meeting presentation slides are available in ADAMSAccession Numbers ML17038A135 and ML17038A136.)
3.Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure VesselNeutron Fluence," U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research,March 2001.
Westinghouse Non-Proprietary Class 3LTR-REA-17-56, Rev. 1Page 44 of 44 April 25, 2017 4.Westinghouse Report WCAP-18060-NP, Rev. 1, "Response to RAIs Concerning the Use ofRAPTOR-M3G for the Catawba Unit 1 Measurement Uncertainty Recapture (MUR) PowerUprate Fluence Evaluations," November 2015. (Available as ADAMS Accession NumberML15324A084.)
 
to W3F1-2017-0027
 
Page 1 of 2 The following changes are made to the LRA as discussed below. Additions are shown with underline and deletions with strikethrough. 
 
The following LRA change corrects a typographical error in the math symbol in
:  XI.E3 Inaccessible Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Non-EQ Inaccessible Power Cables
(  400 V) ( 400V) [B.1.24]
The following changes are being made to the description of the ultimate heat sink to replace the impact of a tornado directly on the wet cooling tower basin with the phrase, "following a tornado,"
as this more accurately reflects the calculation for required makeup to the cooling tower basin.
 
From Normal makeup to the wet cooling tower basins is demineralized water. Replenishment of the wet cooling tower basins from on-site water sources and/or the Mississippi River may be required in response to a tornado event if the tornado ejects water from the cooling tower basins. If available, additional makeup for the wet cooling tower basins can be supplied from onsite storage tanks (condensate storage tank, fire water storage tanks, demineralized water storage tank, and primary water storage tank). Provisions are in place (hose connections) to draw water from these sources via a nonsafety-related, portable, diesel-driven pump, which will supply water via hose directly to the basin. Additional makeup can also be provided directly to the wet cooling tower basins using the portable pump and a fire hose from the potable water supply through a fire hydrant. However, because these sources of water may not be available after a tornado, this function is not an intended function for the systems that include these sources. In the event these sources are not available, provisions are in place to use the portable, diesel-driven pump to supply water from the Mississippi River to the circulating water system, which has piping that can be used to gravity feed makeup to the basin. The function of providing makeup water to the wet cooling tower basin is an intended function of the circulating water system.
From footnote to table of exceptions:
1 The carbon steel main circulating water pipe is in service during normal plant operation. Beginning at the discharge of the circulating water pumps, the 11-foot diameter pipe is internally coated with coal tar epoxy. The pipe connects with an 11-foot diameter uncoated concrete main circulating water pipe on the plant side of the levee. This pipe is in the scope of license renewal to provide a source of water to refill the wet cooling tower basin in the unlikely event that the basin is emptied by following a tornado. A one-time inspection of the coated carbon steel pipe segment is adequate for the following reasons. The probability of a tornado emptying the wet cooling tower striking the Waterford site is extremely low. In the event that any coating material became detached during normal operation, the circulating water flow would transport the coal tar epoxy to the condenser water box. Coating degradation sufficient to cause flow blockage during normal operation would be indicated in the control room as a change in the temperature across any affected water box and as a decrease in condenser vacuum. to W3F1-2017-0027
 
Page 2 of 2 The water boxes are routinely cleaned (approximately once per year) and any degraded coating transported to the water box would be observed during the cleaning operation. A review of WF3 operating experience found no record of degraded circulating water pipe coating plugging condenser tubes. During the tornado event, it is assumed that there is a loss of offsite power resulting in a loss of the main circulating water pumps. Therefore, any detached coal tar epoxy would settle to the bottom of the 11-foot concrete pipe because of the low flow in the pipe and the fact that the coal tar epoxy density is greater than raw water. The 16-inch uncoated water supply line to the wet cooling tower basin attaches to the 11-foot concrete pipe at a point 5.5 feet above the bottom. Thus, any detached coating material would not plug the 16-inch uncoated water supply line. The one-time inspection will be performed within the 10 years prior to the period of extended operation and will include 50 percent of the total length of the pipe segment that is coated with coal tar epoxy as recommended in the NUREG-1801, XI.M42 aging management program described in LR-ISG-2013-01. The acceptance criteria and corrective actions for this inspection will be consistent with the recommendations of NUREG-1801, XI.M42.}}

Latest revision as of 19:51, 17 August 2019

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