CNL-20-008, Application to Modify Watts Bar Nuclear Plant (WBN) Unit 2 Technical Specification 5.9.6, Reactor Coolant System Pressure and Temperature Limits Report, (WBN-TS-19-28)

Application to Modify Watts Bar Nuclear Plant (WBN) Unit 2 Technical Specification 5.9.6, Reactor Coolant System Pressure and Temperature Limits Report, (WBN-TS-19-28)

ML20209A071
Person / Time
Site:	Watts Bar
Issue date:	07/27/2020
From:	Jim Barstow Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CNL-20-008, WBN-TS-19-28
Download: ML20209A071 (32)

Text

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-20-008 July 27, 2020 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 2 Facility Operating License No. NPF-96 NRC Docket No. 50-391

Subject:

Application to Modify Watts Bar Nuclear Plant (WBN) Unit 2 Technical Specification 5.9.6, Reactor Coolant System Pressure and Temperature Limits Report," (WBN-TS-19-28)

In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.90, Application for amendment of license, construction permit, or early site permit, Tennessee Valley Authority (TVA) is submitting a request for an amendment to Facility Operating License No. NPF-96 for the Watts Bar Nuclear Plant (WBN) Unit 2.

The proposed license amendment request (LAR) revises WBN, Unit 2 Technical Specification (TS) 5.9.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR),"

to add WCAP-18124-NP-A Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, as the neutron fluence calculational methodology for the evaluation of reactor vessel specimens, to support determination of RCS pressure and temperature limits.

The enclosure to this submittal provides a description and technical evaluation of the proposed change, a regulatory evaluation, and a discussion of environmental considerations. to the enclosure provides the existing WBN, Unit 2 TS page marked-up to show the proposed changes. Attachment 2 to the enclosure provides the existing WBN, Unit 2 TS page retyped to show the proposed change. Attachment 3 to the enclosure provides the existing WBN, Unit 2 TS Bases pages marked-up to show the proposed changes. Changes to the existing TS Bases are provided for information only and will be implemented under the TS Bases Control Program.

TVA has determined that there are no significant hazard considerations associated with the proposed change and that the change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). In accordance with 10 CFR 50.91, Notice for Public Comment; State Consultation, TVA is sending a copy of this letter and the enclosure to the Tennessee Department of Environment and Conservation.

U.S. Nuclear Regulatory Commission CNL-20-008 Page 2 July 27, 2020 TVA requests approval of the proposed license amendment within one year from the date of this submittal with implementation within 30 days following NRC approval.

There are no new regulatory commitments associated with this submittal. If you have any questions about this proposed change, please contact Gordon Williams, Senior Manager, Fleet Licensing (Acting) at (423) 751-2687.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 27th day of July 2020.

Respectfully, James Barstow Vice President, Nuclear Regulatory Affairs & Support Services

Enclosure:

Evaluation of Proposed Change cc (Enclosure):

NRC Regional Administrator - Region II NRC Project Manager - Watts Bar Nuclear Plant NRC Senior Resident Inspector - Watts Bar Nuclear Plant Director, Division of Radiological Health - Tennessee State Department of Environment and Conservation

Evaluation of Proposed Change

Subject:

Application to Modify Watts Bar Nuclear Plant (WBN) Unit 2 Technical Specification 5.9.6, Reactor Coolant System Pressure and Temperature Limits Report (PTLR)," (WBN-TS-19-28)

CONTENTS 1.0

SUMMARY

DESCRIPTION ............................................................................................. 2 2.0 DETAILED DESCRIPTION.............................................................................................. 2 2.1 Proposed Change ........................................................................................................ 2 2.2 Reason for the Proposed Change ................................................................................ 2

3.0 TECHNICAL EVALUATION

............................................................................................. 3 3.1 Determination of RCS Pressure and Temperature (P/T) Limits .................................... 3 3.2 Traditional Beltline and Non-Traditional (Extended) Beltline......................................... 3 3.3 Evaluation of Proposed TS Changes ........................................................................... 5 3.3.1 Limitation #1 ......................................................................................................... 6 3.3.2 Limitation #2 ........................................................................................................18 3.4 Conclusion ..................................................................................................................19

4.0 REGULATORY EVALUATION

.......................................................................................19 4.1 Applicable Regulatory Requirements and Criteria .......................................................19 4.2 Precedent ...................................................................................................................20 4.3 No Significant Hazards Consideration .........................................................................21 4.4 Conclusion ..................................................................................................................22

5.0 ENVIRONMENTAL CONSIDERATION

..........................................................................22

6.0 REFERENCES

...............................................................................................................22 ATTACHMENTS

1. Proposed TS Changes (Mark-Ups) for WBN Unit 2

2. Proposed TS Changes (Final Typed) for WBN Unit 2

3. Proposed TS Bases Page Changes (Mark-Ups) for WBN Unit 2 (For Information Only)

CNL-20-008 E1 of 23

Enclosure 1.0

SUMMARY

DESCRIPTION In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.90, "Application for amendment of license, construction permit, or early site permit," Tennessee Valley Authority (TVA) is requesting a license amendment to Facility Operating License No. NPF-96 for the Watts Bar Nuclear Plant (WBN), Unit 2. The proposed license amendment request (LAR) revises WBN, Unit 2 Technical Specification (TS) 5.9.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), to add WCAP-18124-NP-A Revision 0, Fluence Determination with RAPTOR-M3G and FERRET (Reference 1), as the neutron fluence calculational methodology for the evaluation of reactor vessel specimens, to support determination of RCS pressure and temperature limits.

2.0 DETAILED DESCRIPTION

2.1 PROPOSED CHANGE

The following is a detailed description of the proposed WBN, Unit 2 TS change.

• TS 5.9.6b is revised to add new item 2 and current TS 5.9.6b.2 is renumbered to 5.9.6b.3 as follows:

2. WCAP-18124-NP-A, Rev. 0 Fluence Determination with RAPTOR-M3G and FERRET may be used as an alternative to Section 2.2 of WCAP-14040-A, Rev. 4.

3. The PTLR will contain the complete identification for each of the TS reference Topical Reports used to prepare the PTLR (i.e., report number, title, revision, date, and any supplements).

• TS Bases B3.4.3, RCS P/T Limits, is revised to add a reference to WCAP-18124-NP-A, Rev. 0.

Both methods have been accepted for use by NRC.

Attachment 1 to the enclosure provides the existing WBN, Unit 2 TS page marked-up to show the proposed changes. Attachment 2 to the enclosure provides the existing WBN, Unit 2 TS page retyped to show the proposed changes. Attachment 3 to the enclosure provides the existing WBN, Unit 2 TS Bases pages marked-up to show the proposed changes. Changes to the existing TS Bases are provided for information only and will be implemented under the TS Bases Control Program.

2.2 REASON FOR THE PROPOSED CHANGE As described in Reference 2, the reactor pressure vessel (RPV) neutron fluence values for WBN, Unit 2 Capsule U were determined using the RAPTOR-M3G computer code described in Reference 1. Therefore, WBN, Unit 2 TS 5.9.6b is being revised to add Reference 1.

This will incorporate a methodology that has been found acceptable by NRC in a safety evaluation for use in calculating neutron fluence. This applies a parallel-processing computer code that allows the evaluation to be completed more quickly without the use of overly conservative simplifying assumptions.

CNL-20-008 E2 of 23

Enclosure

3.0 TECHNICAL EVALUATION

3.1 DETERMINATION OF RCS PRESSURE AND TEMPERATURE LIMITS 10 CFR 50, Appendix G, requires the establishment of Pressure and Temperature Limits (P/T) for specific material fracture toughness requirements of the reactor coolant pressure boundary (RCPB) materials. The reactor vessel beltline is defined in Appendix G as the region of the reactor vessel that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage. The beltline region experiences increased embrittlement over the operating period of the reactor vessel because of accumulated neutron radiation from the core.

Nuclear Regulatory Commission (NRC) Regulatory Issue Summary (RIS) 2014-11 clarifies that P/T limit calculations for ferritic RPV materials, other than those materials with the highest reference temperature, may result in more limiting P/T curves because of higher stresses due to structural discontinuities, such as those in RPV inlet and outlet nozzles. Specifically, all ferritic components within the entire reactor vessel must be considered in the development of P/T limits, and the effects of neutron radiation must be considered for any locations predicted to experience a neutron fluence exposure greater than 1 x 1017 n/cm2 (neutrons per square centimeter) (E > 1 mega-electron volt [MeV]) at the end of the licensed operating period (EOL), in accordance with 10 CFR 50 Appendix G.

The Appendix G requirements include an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. The neutron embrittlement effect on the material toughness is reflected by increasing the nil ductility reference temperature (RTNDT) as exposure to neutron fluence increases. The actual shift in the RTNDT of the RPV material is established periodically by removing and evaluating irradiated reactor vessel material specimens, in accordance with 10 CFR 50, Appendix H. The operating P/T limit curves are adjusted, as necessary, based on the evaluation findings and the recommendations of Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials.

3.2 TRADITIONAL BELTLINE AND NON-TRADITIONAL BELTLINE The traditional beltline region of the RPV encompasses the regions adjacent to and directly surrounding the effective height of the active core. The active core height for WBN, Unit 2 extends from -182.88 cm to 182.88 cm, which represents the axial extension of the traditional beltline region of the reactor vessel. The RPV materials for WBN, Unit 2 are listed in Table 3.2-1.

When compared to the axial elevations of the RPV materials evaluated in Table 3.2-1, the intermediate shell forging 05, lower shell forging 04, and intermediate to lower shell circumferential weld W05 are within the traditional beltline region. These materials have been approved by the NRC for generic application of RAPTOR-M3G for fast neutron fluence determination.

CNL-20-008 E3 of 23

Enclosure Table 3.2-1: Reactor Pressure Vessel Material Locations Distance from Location Midplane of Active Fuel (cm)

Flange Mating Surface 556.823 Centerline of Inlet and Outlet Nozzles 343.463 Lowest Extent of Outlet Nozzle to Shell Weld 276.263 Lowest Extent of Inlet Nozzle to Shell Weld 265.263 Center of Intermediate Shell 05 to Nozzle Shell 06 Circumferential Weld 226.423 Bottom of Upper Core Plate 216.463 Top of Active Fuel 182.88 Center of Lower Shell 04 to Intermediate Shell 05 Circumferential Weld 12.023 Midplane of Active Fuel 0 Bottom of Active Fuel -182.88 Top of Lower Core Plate -191.206 Center of Lower Shell 04 to Bottom Head Ring 03 Circumferential Weld -201.677 Center of Bottom Head Peel 02 to Bottom Head Ring 03 Circumferential Weld -314.777 Lowest extent of r,z model -363.296 Lower Head to Lower Ring Circumferential Weld -401.977 Neutron exposure data for selected pressure vessel materials are provided in Table 3.2-2 for fast neutron (E > 1.0 MeV) fluence. Materials in this table with exposure greater than 1 x 1017 n/cm2 are referred to as part of the non-traditional or extended beltline region and must be evaluated to ensure that the applicable acceptance criteria are met.

CNL-20-008 E4 of 23

Enclosure Table 3.2-2: Fast Neutron Fluence for Selected RPV Materials Maximum Neutron (E > 1.0 MeV) Fluence (n/cm2)

Cumulative Bottom Head Peel 02 Bottom Head Ring 03 Lower Shell 04 Operating to to to Time Bottom Head Ring 03 Lower Shell 04 Int. Shell 05 Cycle (EFPY) Circ. Weld Circ. Weld(a) Lower Shell 04 Circ. Weld 1 0.74 1.33E+14 7.29E+16 5.40E+17 5.11E+17 2 2 3.23E+14 1.77E+17 1.28E+18 1.22E+18 Future 32 4.93E+15 2.47E+18 1.94E+19 1.83E+19 Future 36 5.55E+15 2.78E+18 2.19E+19 2.06E+19 Maximum Neutron (E > 1.0 MeV) Fluence (n/cm2)

Inlet Nozzle to Outlet Nozzle Cumulative Int. Shell 05 Upper Shell 06 to Operating to Weld Upper Shell 06 Time Upper Shell 06 (Lowest Weld Cycle (EFPY) Int. Shell 05 Circ. Weld(b) Extent) (Lowest Extent) 1 0.74 5.12E+17 1.28E+16 5.64E+14 2.85E+14 2 2 1.23E+18 3.28E+16 1.50E+15 7.06E+14 Future 32 1.86E+19 5.12E+17 2.40E+16 1.16E+16 Future 36 2.10E+19 5.75E+17 2.70E+16 1.31E+16 Notes:

(a) The Bottom Head Ring 03 to Lower Shell 04 Circumferential Weld exposure value is representative of the maximum exposure to the Bottom Head Ring 03.

(b) The Intermediate Shell 05 to Upper Shell 06 Circumferential Weld exposure value is representative of the maximum exposure to Upper Shell 06.

3.3 EVALUATION OF PROPOSED TS CHANGES The proposed change to WBN, Unit 2 TS 5.9.6 adds WCAP-18124-NP-A Revision 0, Fluence Determination with RAPTOR-M3G and FERRET as the neutron fluence calculational methodology for the evaluation of reactor vessel specimens. This methodology has been found acceptable by NRC for use in calculating RPV neutron fluence (Reference 3), provided that the limitations and conditions listed in Section 4.0 of Reference 3 are met. Section 4.0 of Reference 3 listed two limitations as discussed in Sections 3.3.1 and 3.3.2 to this enclosure. In addition, RIS 2014-11 identifies materials that should also be considered in the definition of the beltline for development of P/T limits.

CNL-20-008 E5 of 23

Enclosure 3.3.1 Limitation #1 Applicability of WCAP-18124-NP, Revision 0, is limited to the RPV region near the active height of the core based on the uncertainty analysis performed and the measurement data provided. Additional justification should be provided via additional benchmarking, fluence sensitivity analysis to response parameters of interest (e.g., pressure-temperature limits, material stress/strain), margin assessment, or a combination thereof, for applications of the method to components including, but not limited to, the RPV upper circumferential weld and reactor coolant system inlet and outlet nozzles and reactor vessel internal components.

TVA Response In accordance with the above limitation, additional benchmarking is discussed in Section 3.3.1.1 of this enclosure. Margin assessment is discussed in Section 3.3.1.2 of this enclosure. Fluence sensitivity analysis was not used to address this limitation.

3.3.1.1 Additional Benchmarking Measurements In order to collect measurement benchmark data for the extended beltline region, ex-vessel neutron dosimetry (EVND) has been installed at the elevation of the reactor vessel support for a Westinghouse 4-loop plant. The elevation of the reactor vessel support is approximately 8.5 ft above the core midplane. The specific axial locations of the EVND capsules to the core midplane (Z = 0.0 cm) and the time of irradiation are listed in Table 3.3.1.1-1. The dosimeter foil included in the EVND capsules is listed in Table 3.3.1.1-2. The measured dosimetry reactions for those foils are listed in Table 3.3.1.1-3.

Table 3.3.1.1-1: Locations and Time of Irradiation for Sensor Sets Analyzed at RPV Supports Capsule ID Sensor Azimuthal Axial Cycle(s) of Location Location Elevation (cm) Irradiation E RPV Supports 180° 257.99 11 A RPV Supports 225° 255.75 11 K RPV Supports 180° 257.99 12-19 Table 3.3.1.1-2: Foil Sensor Set Contents in EVND at RPV Supports Capsule Radiometric Monitor Foils ID Fe Ni Cu Ti Co Nb U-238 Np-237 E X X X X X X X A X X X X X X X K X X X X X X CNL-20-008 E6 of 23

Enclosure Table 3.3.1.1-3: Measured Dosimetry Reactions in EVND at RPV Supports Material Reaction of Neutron Energy Response1 Product Half-Interest life2 Copper 63 Cu (n,) 60Co 4.53-11.0 MeV 5.271 years Titanium 46 Ti (n,p) 46Sc 3.70-9.43 MeV 83.788 days Iron 54 Fe (n,p) 54Mn 2.27-7.54 MeV 312.13 days Nickel 58 Ni (n,p) 58Co 1.98-7.51 MeV 70.86 days U-238 238 U (n,f) 137Cs 1.44-6.69 MeV 30.07 years Niobium 93 Nb (n,n)93mNb 0.95-5.79 MeV 16.13 years Np-237 237 Np (n,f) 137Cs 0.68-5.61 MeV 30.07 years Cobalt-Al 59 Co (n,) 60Co Thermal 5.271 years Notes:

(1) Energies between which 90% of activity is produced (235U fission spectrum).

Reference ASTM E844-18.

(2) Half-life data is from ASTM E1005-16.

Additional Benchmarking Neutron Transport Calculations The Westinghouse fluence methodology is described in Reference 1. In the application of this methodology to the fast neutron exposure evaluations for the Westinghouse four-loop commercial dosimetry sets, forward transport calculations were carried out to directly solve for the space- and energy-dependent scalar flux, (r,,z,E).

For this additional benchmarking analysis, the transport calculations were carried out using the RAPTOR-M3G three-dimensional discrete ordinates code and the BUGLE-96 (Reference 4) cross-section library. The BUGLE-96 library provides a 67-group coupled neutron-gamma ray group cross-section data set produced specifically for Light Water Reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P3 Legendre expansion and the angular discretization was modeled with an S20 order of angular quadrature.

A plan view of the reactor model is shown in Figure 3.3.1.1-1. In addition to the core, reactor internals, RPV, and concrete bioshield, the model also included explicit representations of the surveillance capsules, RPV clad as well as the RPV nozzles, and supports. Section views of the reactor model are shown in Figure 3.3.1.1-2 and Figure 3.3.1.1-3.

In developing the model of the reactor geometry, nominal design dimensions were used for the various structural components. Water temperatures (and densities) in the core, bypass, and downcomer regions of the reactor were taken to be representative of full-power operating conditions. These coolant temperatures were varied on a cycle-specific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures (e.g., fuel assembly grids, guide tubes).

CNL-20-008 E7 of 23

Enclosure The r,,z geometric mesh description of the reactor model consisted of 241 radial by 190 azimuthal by 469 axial mesh intervals. Mesh sizes were chosen to ensure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion used in the calculations was 0.001.

The core power distributions used in the plant-specific transport analysis included fuel assembly-specific initial enrichments, burnups, and axial power distributions. This information was used to develop spatial- and energy-dependent core source distributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport calculations provided data in terms of the fuel cycle-averaged neutron fluence rate, which when multiplied by the appropriate fuel cycle length, provide the incremental fast neutron fluence exposure for each fuel cycle. The energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial 235U enrichment and burnup history of the individual fuel assemblies.

From the assembly-dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined. These fuel-assembly-specific neutron source strengths derived from the detailed isotopics were then converted from fuel pin Cartesian coordinates to the r,,z spatial mesh arrays used in the RAPTOR-M3G discrete ordinates calculations.

CNL-20-008 E8 of 23

Enclosure Figure 3.3.1.1-1: Reactor Geometry - Plan View at Core Midplane CNL-20-008 E9 of 23

Enclosure Figure 3.3.1.1-2: Reactor Geometry - Section View at Outlet Nozzle Centerline CNL-20-008 E10 of 23

Enclosure Figure 3.3.1.1-3: Reactor Geometry - Section View at Inlet Nozzle Centerline CNL-20-008 E11 of 23

Enclosure Additional Benchmarking Dosimetry Evaluations Evaluations of neutron sensor sets contained in the commercial dosimetry were completed using current state-of-the-art least-squares-methodology described in Section 3.0 of Reference 1.

Least-squares adjustment methods provide the capability of combining the measurement data with the neutron transport calculation resulting in a best-estimate neutron energy spectrum with associated uncertainties. Best-estimates for key exposure parameters, such as fast neutron (E > 1.0 MeV) fluence rate and iron atom displacement rate along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least-squares methods, as applied to reactor dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross sections, and the calculated neutron energy spectrum within their respective uncertainties.

For example, the following equation relates a set of measured reaction rates, Ri, to a single neutron spectrum, g, through the multigroup dosimeter reaction cross section, ig, each with an uncertainty :

The primary objective of the least-squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.

For the least-squares evaluation of the dosimetry, the FERRET code (Reference 5) was used to combine the results of the plant-specific neutron transport calculations and sensor set reaction rate measurements to determine best-estimate values of exposure parameters along with associated uncertainties.

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 rate and associated uncertainty for each sensor contained in the multiple foil set.

3. The energy-dependent dosimetry reaction cross sections and associated uncertainties for each sensor contained in the multiple foil sensor set.

For the current application, the calculated neutron spectrum at each measurement location was obtained from the results of the previously described additional benchmarking neutron transport calculations. The spectrum at each sensor set location was input in an absolute sense (rather than simply a relative spectral shape). Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements. The sensor reaction rates were derived from the measured specific activities of each sensor set and the operating history of the respective fuel cycles. The dosimetry reaction cross sections were obtained from the SNLRML dosimetry cross section library (Reference 6).

CNL-20-008 E12 of 23

Enclosure In addition to the magnitude of the calculated neutron spectra, the measured sensor set reaction rates, and the dosimeter set reaction cross sections, the least-squares procedure requires uncertainty estimates for each of these input parameters. The following provides a summary of the uncertainties associated with the least-squares evaluation of the dosimetry.

Additional Benchmarking Reaction Rate Uncertainties The overall uncertainty associated with the measured reaction rates includes components due to the basic measurement process, the irradiation history corrections, and the corrections for competing reactions. A high level of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures that conform to the ASTM International consensus standards for reaction rate determinations for each sensor type.

After combining all of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures were assigned the following net uncertainties for input to the least-squares evaluation. These uncertainties are given at the 1 level.

Table 3.3.1.1-4: Reaction Rate Uncertainties Reaction Uncertainty 63 Cu (n,) 60Co 5%

46 Ti (n,p) 46Sc 5%

54 Fe (n,p) 54Mn 5%

58 Ni (n,p) 58Co 5%

238 U (n,f) 137Cs 10%

93 Nb (n,n)93mNb 10%

237 Np (n,f) 137Cs 10%

59 Co (n,) 60Co 35% (Note 1)

Note 1: The cobalt content of older Co-Al foils used in EVND is not known for certain, but is believed to be between 0.438% and 0.562%. To account for this unknown, the uncertainty assigned in the least-squares evaluations (typically 5%) was increased by 28% (i.e., 0.562/0.438), which equals 33%, or conservatively rounded to an uncertainty of 35%.

Additional Benchmarking Dosimetry Cross Section Uncertainties As previously noted, the reaction rate cross sections used in the least-squares evaluations were taken from the SNLRML library in Reference 6. This data library provides reaction cross sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross sections and uncertainties are provided in a fine multi-group structure for use in least-squares adjustment applications.

These cross sections were compiled from the ENDF/B-VI cross section evaluations and have been tested with respect to their accuracy and consistency for least-squares evaluations. Further, the library has been empirically tested for use in fission spectra determination as well as in the fluence and energy characterization of 14 MeV neutron sources. Detailed discussions of the contents of the SNLRML library along with the evaluation process for each of the sensors is provided in Reference 6.

CNL-20-008 E13 of 23

Enclosure For sensors included in the dosimetry sets, the following uncertainties in the fission spectrum-averaged cross sections are provided in the SNLRML documentation package.

Table 3.3.1.1-5: Cross Section Uncertainties Reaction Uncertainty 63 Cu (n,) 60Co 4.08-4.16%

46 Ti (n,p) 46Sc 4.51-4.87%

54 Fe (n,p) 54Mn 3.05-3.11%

58 Ni (n,p) 58Co 4.49-4.56%

238 U (n,f) 137Cs 0.54-0.64%

93 Nb (n,n)93mNb 6.96-7.23%

237 Np (n,f) 137Cs 10.32-10.97%

59 Co (n,) 60Co 0.76-3.59%

The ranges shown in Table 3.3.1.1-5 provide an indication of the dosimetry cross section uncertainties associated with the sensor sets used in LWR irradiations.

Additional Benchmarking Calculated Neutron Spectrum Uncertainties While the uncertainties associated with the reaction rates were obtained from the measurement procedures and counting benchmarks, and the dosimetry cross section uncertainties were supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum was constructed from the following relationship:

Where Mgg represents the uncertainty matrix, Rn specifies an overall fractional normalization uncertainty and the fractional uncertainties Rg and Rg specify additional random groupwise uncertainties that are correlated with a correlation matrix given by this correlation matrix equation:

Where:

Pgg represents the correlation matrix. The first term in the correlation matrix equation specifies purely random uncertainties; while the second term describes the short-range correlations over a group range ( specifies the strength of the latter term). The value of is 1.0 when g = g and 0.0 otherwise.

CNL-20-008 E14 of 23

Enclosure The set of parameters defining the input covariance matrix for calculated spectra was as follows:

Flux Normalization Uncertainty (Rn) 15%

Flux Group Uncertainties (Rg, Rg)

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 25%

(E < 0.68 eV) 50%

Short-Range Correlation ()

(E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 Flux Group Correlation Range ()

(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 These uncertainty assignments are consistent with an industry consensus uncertainty of 15-20% (1) for the fast neutron portion of the spectrum and provide for a reasonable increase in the uncertainty for neutrons in the intermediate and thermal energy ranges.

Additional Benchmarking Measurement-to-Calculation Comparison The comparison of the measurement results from each of the sensor set irradiations at RPV supports with corresponding analytical predictions at the measurement locations are presented in Tables 3.3.1.1-6 and 3.3.1.1-7. These comparisons are provided on two levels. On the first level, calculations of individual sensor reaction rates are compared directly with the measured data from the counting laboratories. This level of comparison is not impacted by the least-squares evaluations of the sensor sets. On the second level, calculated values of neutron exposure rates in terms of fast neutron (E > 1.0 MeV) fluence rate and iron atom displacement rate are compared with the best estimate exposure rates obtained from the least-squares evaluation.

In Table 3.3.1.1-6, comparisons of measurement-to-calculation (M/C) ratios are listed for the threshold sensors contained in the EVND dosimetry capsules irradiated at RPV supports that are approximately 8.5 ft above the core midplane. For the individual threshold foils, the average M/C ratio ranges from 0.62 to 1.28, with an overall average of 0.78 and an associated standard deviation of 25.5%. In this case, the overall average was based on an equal weighting of each of the sensor types with no adjustments made to account for the spectral coverage of the individual sensors.

In Table 3.3.1.1-7, best-estimate-to-calculation (BE/C) ratios for fast neutron (E > 1.0 MeV) fluence rate and iron atom displacement rate resulting from the least-squares evaluation of the dosimetry sets is provided for the EVND capsules irradiated at the RPV supports, which are approximately 8.5 ft above the core midplane.

CNL-20-008 E15 of 23

Enclosure The average BE/C ratio for the fast neutron (E > 1.0 MeV) is 0.84 with an associated standard deviation of 8.9% and an average BE/C ratio of 0.94 with an associated standard deviation of 11 % for the iron atom displacement (dpa). These BE/C ratios are within the +/- 20% uncertainty at 1- level recommended by Regulatory Guide (RG) 1.190.

For the extended beltline region, the M/C data provided in Table 3.3.1.1-6 and the BE/C data provided in Table 3.3.1-5 suggest that the calculations are over predicting the neutron exposure, in particular at the high end of the energy spectrum. For instance, the bottom of the 90% neutron response for the copper, titanium, iron, and nickel dosimeters is 4.53 MeV, 3.70 MeV, 2.27 MeV, and 1.98 MeV, respectively. Neutrons with energies greater than these constitute a small fraction of the neutron (E > 1.0 MeV) fluence rate in the extended beltline region. The BE/C values in Table 3.3.1.1-7 account for the spectral coverage of the different sensors, and provide an estimate of the key damage parameters, fast neutron (E > 1.0 MeV) fluence rate and iron atom displacement rate, that result from an uncertainty-weighted reconciliation of all of the measurements and calculations. The BE/C values in Table 3.3.1.1-7 suggest that the calculated damage parameters are moderately conservative relative to the best-estimate values.

The uncertainty associated with a fluence determination methodology is comprised of two major components: the results of an analytic uncertainty analysis and the results of benchmarking comparisons. An analytic uncertainty analysis assesses the level of confidence in key input parameters to a fluence calculation and quantifies the impact that plausible input parameter variations have on calculated fluence results.

Benchmarking comparisons refer to comparisons of fluence calculations performed with a candidate methodology to alternate calculations or to measurements from a representative environment.

The fast neutron (E > 1.0 MeV) fluence rate analytic uncertainty for the Westinghouse fluence methodology in the reactor cavity at locations opposite the top and bottom of the active fuel is 17-18% at the 1 level (Reference 1). Combining this analytic uncertainty with the other uncertainty components identified in Reference 1 yields a net uncertainty (1) estimate of 19-20%.

A comprehensive analytic uncertainty assessment for the extended beltline region has not yet been completed. Preliminary estimates suggest that the analytic uncertainty in this region will be modestly greater than the analytic uncertainty attributed to the cavity locations at the elevation of the top of the core. The benchmarking data set for the extended beltline region evaluated in this document is small (three dosimetry capsules);

however, it appears that, after combining the results in this document with the expected results of a comprehensive analytic uncertainty assessment, the net methodology uncertainty for the RAPTOR-M3G methodology may be on the order of 30% in the vicinity of the RPV supports.

CNL-20-008 E16 of 23

Enclosure Table 3.3.1.1-6: Measured-to-Calculated (M/C) Reaction Rates - Ex-Vessel Capsule Located in the Vicinity of the RPV Supports Reaction Capsule E Capsule A Capsule K Average  % Std. Dev.

63 Cu (n,) 60Co 0.65 - 0.58 0.62 8.0 46 Ti (n,p) 46Sc 0.73 0.65 0.67 0.68 6.1 54 Fe (n,p) 54Mn 0.72 0.66 0.64 0.67 6.2 58 Ni (n,p) 58Co 0.75 0.67 0.65 0.69 7.7 238 U (n,f) 137Cs 1.03 0.89 - 0.96 10.3 93 Nb (n,n)93mNb - - 1.28 1.28 -

237 Np (n,f) 137Cs 1.12 0.80 - 0.96 23.6 Average of M/C Results 0.78 25.5 Table 3.3.1.1-7: Best-Estimate-to-Calculated (BE/C) Exposure Rates - Ex-Vessel Capsule Located in the Vicinity of the RPV Supports Capsule Neutron (E > 1.0 MeV) Fluence Rate Iron Atom Displacement Rate BE/C BE/C E 0.91 0.99 A 0.76 0.82 K 0.84 1.00 Average 0.84 0.94

% Std. Dev. 8.9 11 Discussion of the Difference in Treatment of the Order of Angular Quadrature In the most recent fluence analysis for WBN, Unit 2 (Reference 2), anisotropic scattering was treated with a P3 Legendre expansion, and angular discretization was modeled with an S12 order of angular quadrature. In the additional benchmarking discussed above, anisotropic scattering was treated with a P3 Legendre expansion and the angular discretization was modeled with an S20 order of angular quadrature. The maximum difference observed between S12 and S20 calculations, considering both the inside and outside surface of the reactor vessel, is 5%. Considering that this additional uncertainty may be combined in quadrature with the results from additional benchmarking discussed above, the increase in uncertainty is minimal. Therefore, the maximum range discussed above (i.e., 5%) is used. The RAPTOR-M3G fluence determination methodology has about 30% uncertainty in the fast neutron (E > 1.0 MeV) determination.

The RPV extended beltline materials evaluated for WBN, Unit 2 in Table 3.2-2 are mostly located within an axial distance of 8.5 ft above or below the core midplane. The exceptions are the two materials used at the lowest extent of the outlet nozzle to shell weld (about 9 ft above the core midplane) and at the center of bottom head peel 02 to bottom head ring 03 circumferential weld (about 10 ft below the core midplane).

However, the maximum EOL fluence for these two materials was calculated to be 1.31E16 n/cm2, which is more than a factor of five lower (i.e., 500% lower) than the prescribed threshold of 1E17 n/cm2 for the definition of the extended beltline region.

Because the RAPTOR-M3G fluence determination methodology uncertainty only increases from 19-20% at the top of the active fuel (+182.88 cm above core midplane) to approximately 30% at the RPV supports (+257.99 cm above the core midplane), it is not credible that the methodology uncertainty would increase from approximately 30% to CNL-20-008 E17 of 23

Enclosure 500% for an additional 1.5 ft axial distance from the core midplane. Thus, these two materials are not required to be included as extended beltline material and do not require evaluation to ensure that the applicable acceptance criteria are met.

3.3.1.2 Margin Assessment The results from additional benchmarking discussed above can also be used to estimate the uncertainty for the inlet and outlet nozzles that are evaluated for fracture toughness due to structural discontinuities, as discussed in RIS 2014-11. Therefore the methodology uncertainty for fluence determination for these RPV extended beltline materials is also approximately 30% and the calculated fast neutron (E > 1.0 MeV) fluence tends to be over estimated when compared to the measurement benchmark data. Additionally, Reference 7 recognizes that unless the fast neutron (E > 1.0 MeV) fluence for the nozzle material is greater than the Regulatory Guide 1.99, Revision 2 fluence value of 4.28E17 n/cm2, embrittlement need not be considered for nozzle forging evaluation and the nozzles will be non-limiting compared to the beltline with respect to the pressure temperature limit curves. The fast neutron (E > 1.0 MeV) fluence values reported for both the inlet and outlet nozzles in Table 3.2-2 are more than a factor of three lower than the prescribed threshold of 1E17 n/cm2 for the definition of the extended beltline region, and a factor of 15 lower than the threshold defined in Reference 7. As the evaluated net RAPTOR-M3G methodology uncertainty is approximately 30% for this elevation based on additional benchmarking and preliminary analytical uncertainty analysis, it is also not credible that the inlet and outlet nozzle fast neutron (E > 1.0 MeV) fluence at the EOL will exceed 4.28E17 n/cm2. Therefore, there is significant margin prior to these nozzle materials becoming limiting.

Based on the additional benchmarking at the RPV extended beltline region and margin assessment discussed herein, the RAPTOR-M3G fluence determination methodology is justified to be applicable to the WBN, Unit 2 RPV extended beltline region fast neutron (E > 1.0 MeV) fluence determination for fracture toughness evaluation.

Therefore, Limitation #1 has been addressed.

3.3.2 Limitation #2 Least squares adjustment is acceptable if the adjustments to the M/C ratios and to the calculated spectra values are within the assigned uncertainties of the calculated spectra, the dosimetry measured reaction rates, and the dosimetry reaction cross sections.

Should this not be the case, the user should re-examine both measured and calculated values for possible errors. If errors cannot be found, the particular values causing the inconsistency should be disqualified.

TVA Response Limitation #2 applies in situations where the least squares analysis is used to adjust the calculated values of neutron exposure. However, in the most recent capsule analysis (Reference 2) that utilized the methodology described in Reference 1, the least squares analysis is provided only as a supplemental check on the results of the dosimetry evaluation. The least squares analysis was not used to modify the calculated surveillance capsule or reactor pressure vessel neutron exposure. Therefore, Limitation #2 has been addressed.

CNL-20-008 E18 of 23

Enclosure

3.4 CONCLUSION

By properly addressing the limitations set forth in the NRC Safety Evaluation for WCAP-18124-NP-A Revision 0, and by evaluating the materials identified in RIS 2014-11 for consideration, the neutron fluence calculational methodology described in this enclosure has been shown to be acceptable for use at WBN, Unit 2 for the evaluation of reactor vessel specimens.

4.0 REGULATORY EVALUATION

4.1 APPLICABLE REGULATORY REQUIREMENTS AND CRITERIA Regulations 10 CFR 50.60, Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation, imposes fracture toughness and material surveillance program requirements which are set forth in 10 CFR 50, Appendices G, Fracture Toughness Requirements, and H, Reactor Vessel Material Surveillance Program Requirements.

10 CFR 50 Appendix G specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime.

Compliance with Appendix G is described in Section 5.2 of the WBN dual-unit Updated Final Safety Analysis Report (UFSAR).

10 CFR 50 Appendix H requires a program to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of light water nuclear power reactors which result from exposure of these materials to neutron irradiation and the thermal environment. Compliance with Appendix H is described in Section 5.4 of the WBN UFSAR.

General Design Criteria WBN, Unit 2 was designed to meet the intent of the "Proposed General Design Criteria for Nuclear Power Plant Construction Permits" published in July 1967. The WBN construction permit was issued in January 1973. The UFSAR, however, addresses the General Design Criteria (GDC) published as Appendix A to 10 CFR 50 in July 1971.

Conformance with the GDCs is described in Section 3.1.2 of the UFSAR.

GDC 31, Fracture Prevention of Reactor Coolant Pressure Boundary, requires the reactor coolant pressure boundary to be designed with sufficient margin to assure that, when stressed under operating, maintenance, testing, and postulated accident conditions, (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, CNL-20-008 E19 of 23

Enclosure maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws. Compliance with GDC 31 is described in Section 3.1.2.4 of the WBN UFSAR.

GDC 32, Inspection of Reactor Coolant Pressure Boundary, requires that components which are part of the reactor coolant pressure boundary, shall be designed to permit (1) periodic inspection and testing of important areas and features to assess their structural and leaktight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel. Compliance with GDC 32 is described in Section 3.1.2.4 of the WBN UFSAR.

Regulatory Guidance Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," contains guidance on methodologies the NRC considers acceptable for determining the effect of neutron radiation on reactor vessel materials. This RG is used as a reference in WCAP-18124-NP-A (Reference 1).

RG 1.190, Revision 0, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," describes a methodology acceptable to the NRC staff for determining the best-estimate neutron fluence experienced by materials in the beltline region of light water reactor (LWR) pressure vessels, as well as for determining the overall uncertainty associated with those best-estimate values. In Reference 3, NRC stated that the calculational fluence methodology described in the WCAP adheres to the guidance in RG 1.190.

RIS 2014-11 clarifies that P/T limit calculations for ferritic RPV materials other than those materials with the highest reference temperature may result in more limiting P/T curves because of higher stresses due to structural discontinuities, such as those in RPV inlet and outlet nozzles. Conformance with RIS 2014-11 is described in Section 3 of this enclosure.

4.2 PRECEDENT While there is no exact precedent for this LAR, use of the methodology described in Reference 1 for calculating RPV neutron fluence has been found to be acceptable by NRC in the following cases.

• License Amendment Nos. 281 and 277 regarding Measurement Uncertainty Recapture Power Uprate for the Catawba Nuclear Station, Units 1 and 2, respectively, dated April 29, 2016 (ML16081A333). Section 3.2.6.3 of the NRC Safety Evaluation, states the NRC staff concludes that the LAR is acceptable, with respect to the use of the RAPTOR-M3G neutron fluence calculation.

• License Amendment No. 252 for the Waterford Steam Electric Station Unit 3, dated July 23, 2018 (ML18180A298) which revised Section 4.3.3, "Analytical Methods," of the Updated Final Safety Analysis Report to indicate that the RAPTOR-M3G fluence code is used for reactor vessel fluence calculations.

CNL-20-008 E20 of 23

Enclosure 4.3 NO SIGNIFICANT HAZARDS CONSIDERATION Tennessee Valley Authority (TVA) proposes to revise the Watts Bar Nuclear Plant (WBN), Unit 2 Technical Specifications (TS) 5.9.6 Reactor Coolant System (RCS)

Pressure and Temperature Limits Report (PTLR)," to add WCAP-18124-NP-A Revision 0 as the neutron fluence calculational methodology for the evaluation of reactor vessel specimens, to support determination of RCS pressure and temperature limits. This will incorporate a methodology that has been found acceptable by NRC in a safety evaluation for use in calculating neutron fluence. This applies a parallel-processing computer code that allows the evaluation to be completed more quickly without the use of overly conservative simplifying assumptions.

TVA has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequence of an accident previously evaluated?

Response: No The proposed changes do not require physical changes to plant systems, structures, or components. There is no interaction with a potential accident-initiating mechanism.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes do not introduce new or different accidents to be postulated and subsequently evaluated, and no changes are being made to the plant that would introduce any new accident causal mechanisms. This license amendment request does not affect any plant systems that are potential accident initiators; nor does it have any significantly adverse effect on any accident mitigating systems.

The proposed change does not change the functional requirements, configuration, or method of operation of any system or component. Under the proposed change, no additional plant equipment will be installed.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

CNL-20-008 E21 of 23

Enclosure

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes do not alter the permanent plant design, nor does it change the assumptions contained in the safety analyses. No safety limits or operating parameters used to establish the safety margin are affected. The safety margins included in analyses of accidents are not affected by the proposed change.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, TVA concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

4.4 CONCLUSION

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

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. Westinghouse Report, WCAP-18124-NP-A Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, dated July 2018 (ML18204A010)

2. Westinghouse Report, WCAP-18518-NP Revision 0, Analysis of Capsule U from the Watts Bar Unit 2 Reactor Vessel Radiation Surveillance Program, dated March 2020 (ML20107F717)

3. Final Safety Evaluation for Topical Report WCAP-18124-NP Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, Westinghouse Electric Company, dated June 15, 2018 (ML18156A066)

CNL-20-008 E22 of 23

Enclosure

4. 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, March 1996 (Available from the Radiation Safety Information Computational Center, Oak Ridge National Laboratory.)

5. RSICC Computer Code Collection PSR-145 FERRET: Least-Squares Solution to Nuclear Data and Reactor Physics Problems, January 1980 (Available from the Radiation Safety Information Computational Center, Oak Ridge National Laboratory.)

6. RSICC Data Library Collection DLC-178, SNLRML Recommended Dosimetry Cross Section Compendium, July 1994 (Available from the Radiation Safety Information Computational Center, Oak Ridge National Laboratory.)

7. Pressurized Water Reactor Owners Group (PWROG) Report, PWROG-15109-NP-A Revision 0, PWR Pressure Vessel Nozzle Appendix G Evaluation, dated January 2020 (ML20024E573)

CNL-20-008 E23 of 23

Attachment 1 Proposed TS Changes (Mark-Ups) for WBN Unit 2 CNL-20-008

Reporting Requirements 5.9 5.9 Reporting Requirements (continued) 5.9.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation (power operated relief valve lift settings required to support the Cold Overpressure Mitigation System (COMS) and the COMS arming temperature), criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3 RCS Pressure and Temperature (P/T) Limits LCO 3.4.12 Cold Overpressure Mitigation System (COMS)

b. The analytical methods used to determine the RCS pressure and temperature limits and COMS setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. WCAP-14040-A, Rev. 4 Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves.

2. WCAP-18124-NP-A, Rev. 0 Fluence Determination with RAPTOR-M3G and FERRET may be used as an alternative to Section 2.2 of WCAP-14040-A Rev. 4.

3. The PTLR will contain the complete identification for each of the TS reference Topical Reports used to prepare the PTLR (i.e., report number, title, revision, date, and any supplements).

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

Watts Bar - Unit 2 5.0-34 (continued)

Amendment XX

Attachment 2 Proposed TS Changes (Final Typed) for WBN Unit 2 CNL-20-008

Reporting Requirements 5.9 5.9 Reporting Requirements (continued) 5.9.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation (power operated relief valve lift settings required to support the Cold Overpressure Mitigation System (COMS) and the COMS arming temperature), criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3 RCS Pressure and Temperature (P/T) Limits LCO 3.4.12 Cold Overpressure Mitigation System (COMS)

b. The analytical methods used to determine the RCS pressure and temperature limits and COMS setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. WCAP-14040-A, Rev. 4 Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves.

2. WCAP-18124-NP-A, Rev. 0 Fluence Determination with RAPTOR-M3G and FERRET may be used as an alternative to Section 2.2 of WCAP-14040-A Rev. 4.

3. The PTLR will contain the complete identification for each of the TS reference Topical Reports used to prepare the PTLR (i.e., report number, title, revision, date, and any supplements).

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

Watts Bar - Unit 2 5.0-34 (continued)

Amendment XX

Attachment 3 Proposed TS Bases Changes (Mark-Ups) for WBN Unit 2 (For Information Only)

CNL-20-008

RCS P/T Limits B 3.4.3 BASES (continued)

APPLICABLE The P/T limits are not derived from Design Basis Accident (DBA)

SAFETY analyses. They are prescribed during normal operation to avoid ANALYSES encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, an unanalyzed condition. References 8 and 9 establishes the methodology for determining the P/T limits.

Although the P/T limits are not derived from any DBA, the P/T limits are acceptance limits since they preclude operation in an unanalyzed condition.

RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The two elements of this LCO are:

a. The limit curves for heatup, cooldown, and ISLH testing; and

b. Limits on the rate of change of temperature.

The LCO limits apply to all components of the RCS, except the pressurizer. These limits define allowable operating regions and permit a large number of operating cycles while providing a wide margin to nonductile failure.

The limits for the rate of change of temperature control and the thermal gradient through the vessel wall are used as inputs for calculating the heatup, cooldown, and ISLH testing P/T limit curves. Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the P/T limit curves.

Violating the LCO limits places the reactor vessel outside of the bounds of the stress analyses and can increase stresses in other RCPB components. The consequences depend on several factors, as follow:

a. The severity of the departure from the allowable operating P/T regime or the severity of the rate of change of temperature;

b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and

c. The existences, sizes, and orientations of flaws in the vessel material.

Watts Bar - Unit 2 B 3.4-11 (continued)

Amendment XX

RCS P/T Limits B 3.4.3 BASES SURVEILLANCE SR 3.4.3.1 REQUIREMENTS Verification that operation is within the PTLR limits is when RCS pressure and temperature conditions are undergoing planned changes. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.

This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement.

REFERENCES 1. Appendix "B" to RCS System Description N3-68-4001, "Watts Bar Unit 2 RCS Pressure and Temperature Limits Report."

2. Title 10, Code of Federal Regulations, Part 50, Appendix G, "Fracture Toughness Requirements."

3. ASME Boiler and Pressure Vessel Code,Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure."

4. ASTM E 185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels,"

July 1982.

5. Title 10, Code of Federal Regulations, Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements."

6. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.

7. ASME Boiler and Pressure Vessel Code,Section XI, Appendix E, "Evaluation of Unanticipated Operating Events."

8. WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.

9. WCAP-18124-NP-A, Revision 0 Fluence Determination with RAPTOR-M3G and FERRET, July 2018.

Watts Bar - Unit 2 B 3.4-15 Revision 34 Amendment 36 Amendment XX