ML20315A424

Enclosure 7 - CoC 1042 Amendment 2, Revision 6 UFSAR Changed Pages

ML20315A424
Person / Time
Site:	07201042
Issue date:	10/29/2020
From:	Orano TN Americas, TN Americas LLC
To:	Office of Nuclear Material Safety and Safeguards
Shared Package
ML20315A417	List: ML20315A418 ML20315A419 ML20315A420 ML20315A421 ML20315A422 ML20315A424 ML20315A425 ML20315A426 ML20315A427 ML20315A428
References
E-57418, EPID L-2019-LLA-0078
Download: ML20315A424 (53)

Text

Enclosure 7 to E-57418 CoC 1042 Amendment 2, Revision 6 UFSAR Changed Pages (Public Version)

Proprietary and Security Related Information for Drawing EOS01-1000-SAR, Rev. 4A Withheld Pursuant to 10 CFR 2.390

Proprietary and Security Related Information for Drawing EOS01-1005-SAR, Rev. 3A Withheld Pursuant to 10 CFR 2.390

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 2-38 Table 2-5 EOS-37PTH/EOS-89BTH DSC Shell Assembly Loads and Load Combinations Loading Type DSC Orientation Load for Analysis Load Combination Service Level Load Combination No.

Dead Weight (DW)

Horizontal(11) 1 g down DW + Pressure + Seismic (S)

D 10 Internal Pressure - Off-Normal 20 psig EOS-HSM HSM-MX S =

+/- 1.25g (axial)

+/- 2g (transverse)

+/- 1g (vertical)

S =

+/- 1.5g(axial)

+/- 1.6g(transverse

+/- 0.8g(vertical)

Seismic (S)

Test Pressure at fabricator - 18 psig(12)

Vertical See Note 14 Test 11 External pressure Horizontal D

12 Loss of neutron shield/air circulation Horizontal 130 psig Internal Pressure-Accident D

13(18)

Notes

1.

DSC in EOS-TC in vertical orientation. Only inner top cover is installed.

2.

Bounding thermal case for normal operations of TC in vertical orientation.

3.

DSC in EOS-TC; EOS-TC is in horizontal orientation and supported at the trunnion and saddle locations.

4.

Not used.

5.

The push loads are applied at the canister bottom surface within the grapple ring support.

6.

The pull loads are applied at the inner surface of the grapple ring.

7.

Level B evaluations may take credit for 10% increase in allowable per NB-3223(a).

8.

Controlling thermal off-normal case.

9.

Load combination results to bound cases with and without internal pressure. Level B is used for the case with internal pressure. Level A is used for the case without internal pressure.

10. Bounding pressure of EOS-HSM blocked vent accident or TC accident fire conditions.

11. DSC in EOS-HSM supported on the steel rails. DSC in HSM-MX supported on front and rear DSC supports.

12. Conservatively used 18 psig as the test pressure; test configuration is circular shell and inner bottom welded to shell; a top end lid with a 155 kips clamping force used to seal the test assembly.

All Indicated Changes are in response to Revised RAI 8-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 2-39

13. (General) Material properties at temperature.

14. The maximum accident condition external pressure before DSC collapse/buckling is determined.

15. (General) In addition to the Part 72 loads, postulated end, corner, and side drops associated with 10 CFR Part 71 are evaluated to ensure that the DSC can be licensed as a transportable system.

16. These handling loads in conjunction with Level A limits bounds case of EOS-TC in fuel building under seismic loads (Level D accident condition).

17. The top end drop and bottom end drop are not credible events under 10 CFR Part 72; therefore these drop analyses are not required. However, consideration of end drops (for 10 CFR Part 71 conditions) and the 65-inch side drop conservatively envelope the effects of a corner drop.

18. Load case 13 is performed to validate the shell performance after a drop accident at maximum shell temperatures for Duplex stainless steels only since strength properties for the duplex stainless steel deteriorates significantly past 650°F.

All Indicated Changes are in response to Revised RAI 8-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 3.9.1-2 stress intensity level for the combined loads, instead of at component stress level, is also a conservative method for reducing the number of analysis runs.

The stresses of all components are assessed by means of elastic analysis methodology for all load combinations, except for accident loading conditions.

Elastic-plastic analysis methodology is used to assess the stresses for Service Level D load combinations.

A detailed description of each load combination is provided in Section 3.9.1.2.8.

Material Properties 3.9.1.2.1 For elastic analysis, temperature dependent material properties used for each component of DSC shell assembly are obtained from the American Society of Mechanical Engineers (ASME) code [3.9.1-2], and are summarized in Chapter

8. Material properties used for stress evaluations are conservatively taken at 500 °F under normal loading conditions. For the partial penetration welds and grapple assembly, 350 °F allowable stresses are used for comparison to load induced stresses as these components remain below this lower temperature under normal loading conditions.

For plastic analysis, a bilinear stress-strain curve with a 5% tangent modulus is used for steel components. The non-linear material properties at 500 °F for side drop analysis are shown in Table 3.9.1-3. Steel material (except shield plugs) is modeled by bilinear kinematic hardening method (TB, BKIN - [3.9.1-9]).

DSC Shell Stress Criteria 3.9.1.2.2 The calculated stresses in the DSC shell assembly structural components are compared with the allowable stresses set forth by ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Subsection NB [3.9.1-3] under normal (Level A), and off-normal (Level B) loading conditions. Appendix F of the ASME B&PV Code is used to evaluate the calculated stresses in the DSC shell assembly under accident (Level D) loading conditions. Allowable stress limits for Levels A, B and D service loading conditions, as appropriate, are summarized in Table 3-1, and the corresponding allowable stress values at different temperatures are summarized in Table 3.9.1-5.

The OTCP-to-DSC shell weld and the ITCP-to-DSC shell weld, which are both partial penetration welds, are to be evaluated using a joint efficiency factor of 0.8. Per NUREG-1536 [3.9.1-7], the minimum inspection requirement for end closure welds is multi-pass dye penetrant testing (PT) using a stress (allowable) reduction factor of 0.8. The allowable weld stresses are summarized in Table 3.9.1-4.

All Indicated Changes are in response to Revised RAI 8-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 3.9.1-25 Load Combination 8 Load Combination 8 (LC8) addresses the DSC when it is in the horizontal position. LC8 describes the accident internal pressure load case. The stress intensities from the dead weight load case and the 130-psig internal pressure load case are determined independently and subsequently added to develop the maximum stress intensity on the DSC component and weld.

Load Combination 9 Load Combination 9 (LC9) addresses the DSC when it is in the horizontal position. LC9 describes the off-normal internal pressure load case. The stress intensities from the dead weight load case, 20-psig internal pressure load case and thermal load cases are determined independently and subsequently added to develop the maximum stress intensity on the DSC component and weld.

Load Combination 10 Load Combination 10 (LC10) addresses the DSC when it is in the horizontal position. LC10 describes the seismic load case. LC10 is developed by post-processing the stresses from an FEM that includes the internal off-normal pressure loads and the seismic loads.

Load Combination 11 Load Combination 11 (LC11) addresses the DSC when it is in the vertical position. LC11 describes the fabrication pressure and leak testing loads. LC11 is developed by post-processing the stresses from stresses from the fabrication pressure/leak test load case.

Load Combination 13 Load Combination 13 (LC13) addresses the DSC when it is in the horizontal position. LC13 addresses the accident internal pressure load case post-processed at the accident temperature material properties. Load case 13 is performed to validate the shell performance after a drop accident at maximum shell temperatures for duplex stainless steels only since strength properties for the duplex stainless steel deteriorates significantly past 650°F. This is conservatively post-processed using 304SS strength properties DSC Shell Buckling Evaluation 3.9.1.3 An FE plastic analysis with large displacement option is performed to monitor occurrence of canister shell buckling under the specified loads.

The bottom end drop envelopes the top end drop because the top end structure is heavier than the bottom end structure, which will impose a larger load on the DSC shell. A drop on the bottom end is therefore chosen for buckling analysis.

All Indicated Changes are in response to Revised RAI 8-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 3.9.1-30 Flaw Size Calculation 3.9.1.5.2 For 3D, half-symmetric model, as described in Section 3.9.1.2.3, the tensile radial membrane stresses in the weld are evaluated by the stress linearization method explained in Section 3.9.1.2.5.

Radial stresses for controlling load combination are calculated by adding individual load cases. Bounding radial tensile stresses in OTCP weld for all load combinations for Service Level A, B, and D are assessed. The allowable flaw depths, calculated by means of the methodology described in previous Section and are shown in Table 3.9.1-13.

Based on the evaluation, requirements for welding and weld inspections should be based on limiting the weld critical depth for surface and subsurface flaws to the following values:

Surface Crack:

0.38 inch.

Subsurface Crack:

0.38 inch.

Critical Flaw Size of Duplex Stainless Steel at Accident Thermal Temperatures 3.9.1.5.3 A loss of neutron shield and/or loss of air circulation accident is possible after a drop accident. Table 4.9.7-7 reports the maximum steady-state temperatures of various components after these thermal accidents. Standard stainless steels maintain strength properties through 800°F per ASME Section II Part D, therefore the accident condition evaluated in LC13 is sufficient to validate the adequacy of the confinement boundary. [

]

All Indicated Changes are in response to Revised RAI 8-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 3.9.1-31 Conclusions 3.9.1.6 The EOS DSC shell assembly has been analyzed for normal, off-normal, and accident load conditions using three dimensional finite element analyses. The load combinations provided in Section 3.9.1.2.8 are used in the analysis of the EOS DSC. Stress intensities in different components of the DSC shell assembly, compared with ASME code stress intensity allowables and the resulting stress ratios are summarized in Table 3.9.1-7, Table 3.9.1-7a, Table 3.9.1-8, Table 3.9.1-9, Table 3.9.1-10, Table 3.9.1-11, Table 3.9.1-12, and Table 3.9.1-14 for DSCs supported by the DSC support structure and Table 3.9.1-7b, Table 3.9.1-7c, Table 3.9.1-8a, Table 3.9.1-9a, Table 3.9.1-10a, Table 3.9.1-11a, and Table 3.9.1-12a for DSCs supported by the FPS DSC support structure. FPS DSC support structure evaluations provide results for the maximum and the minimum lengths of canisters for the EOS-HSM-FPS and EOS-HSMS-FPS designs. Minimum length DSC is determined to be the controlling configuration in DSC evaluations. Results obtained for minimum canister length are documented in this chapter. The stress ratio is calculated by dividing the maximum stress intensity by the stress intensity allowable value, with the stress ratio required to be less than 1. Figure 3.9.1-20 shows the linearized component stresses for the DSC shell for the internal pressure (normal) load case. Figure 3.9.1-27 shows the strain criteria state of the DSC.

All Indicated Changes are in response to Revised RAI 8-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 3.9.1-33 References 3.9.1.7 Title 10, Code of Federal Regulations, Part 72, Licensing Requirements for the 3.9.1-1 Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater than Class C Waste.

American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel 3.9.1-2 Code,Section II, Part D, 2010 Edition through 2011 Addenda.

American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel 3.9.1-3 Code,Section III, 2010 Edition through 2011 Addenda.

American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel 3.9.1-4 Code,Section XI, Division 1, Appendix C, 2010 Edition Addenda through 2011 Addenda.

ANSI N14.5, Leakage Tests on Packages for Shipment of Radioactive 3.9.1-5 Materials, 1997.

ANSI N14.6 - 1993, American National Standard for Radioactive Materials -

3.9.1-6 Special Lifting Devices for Shipping Containers Weighing 10000 pounds (4500 kg) or More, American National Standards Institute, Inc., New York.

NUREG-1536, Standard Review Plan for Spent Fuel Dry Cask Storage 3.9.1-7 Systems at a General License Facility, Revision 1, U.S. Nuclear Regulatory Commission, July 2010.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.60, Design 3.9.1-8 Response Spectra for Seismic Design of Nuclear Power Plants, Revision 1, 1973.

ANSYS Computer Code and Users Manual, Release 14.0, 14.0.3, and 17.1.

3.9.1-9 American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel 3.9.1-10 Code, 2013,Section III Appendices.

NUREG CR/1815, Recommendations for Protecting Against Failure by Brittle 3.9.1-11 Fracture in Ferritic Steel Shipping Containers Up to Four Inches Thick.

475 °C Embrittlement in a Duplex Stainless Steel UNS S31803, Materials 3.9.1-12 Research, Vol. 4, No. 4, 237-240, 2001.

Welding Research Council Bulletin 265, Interpretive Report on Small Scale 3.9.1-13 Test Correlations with KIc Data, February 1981.

All Indicated Changes are in response to Revised RAI 8-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 3.9.1-36 Table 3.9.1-4 Allowable Weld Stresses for Pressure Boundary Partial Penetration Welds, Material Type 304 Service Level Stress Region / Category Stress Criteria Allowable Stress Value at 350 °F [ksi]

Level A /

Level B Primary Local Membrane Stress, PL PL = 0.8 [1.5 Sm]

23.2 Primary Local Membrane +

Bending Stress, PL + Pb PL + Pb = 0.8 [1.5 Sm]

23.2 Primary + Secondary Stress, P+Q PL + Pb + Q = 0.8 [3.0 Sm]

46.3 Level D (Elastic)

Primary Local Membrane Stress, PL 0.8 [Min(3.6 Sm, Su)]

52.08 Primary Local Membrane +

Bending Stress, PL + Pb 0.8 [Min(3.6 Sm, Su)]

52.08 Level D (Elastic Plastic)

Primary Local Membrane Stress, PL 0.8 [0.9 Su]

46.9 Primary Local Membrane +

Bending Stress, PL + Pb 0.8 [0.9 Su]

46.9 Table 3.9.1-5 SA-240/SA-479 304 & SA-182 F304 -Stress Allowables Temp

(°F)

Sm (ksi)

Sy (ksi)

Su (ksi)

Level A/B Level D (Elastic) Level D (Plastic)

Pm Pm +

Pb Pm +

Pb + Q Pm Pm +

Pb Pm Pm +

Pb 70 20 30 75 20.0 30.0 60.0 48.0 72.0 52.5 67.5 200 20 25 71 20.0 30.0 60.0 48.0 71.0 49.7 63.9 300 20 22.4 66.2 20.0 30.0 60.0 46.3 66.2 46.3 59.6 400 18.6 20.7 64 18.6 27.9 55.8 44.6 64.0 44.8 57.6 500 17.5 19.4 63.4 17.5 26.3 52.5 42.0 63.0 44.4 57.1 600 16.6 18.4 63.4 16.6 24.9 49.8 39.8 59.8 44.4 57.1 700 15.8 17.6 63.4 15.8 23.7 47.4 37.9 56.9 44.4 57.1 750 15.5 17.2 63.3 15.5 23.3 46.5 37.2 55.8 44.3 57.0 All Indicated Changes are in response to Revised RAI 8-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 3.9.1-40 Table 3.9.1-7 DSC Shell Stress Results, Confinement Boundary - Load Combinations 3 Pages Load Comb No.

Service Level DSC Orientation Stress Category Loads Stress intensity (ksi)

Allowable Stress (ksi)

Stress Ratio 10 D

Horizontal(3)

Pm DWh + max.(HS_TOP, HS_BOT)+PI(20) 9.25 44.38 0.21 Pm+Pb DWh + max.(HS_TOP, HS_BOT)+PI(20) 28.98 57.06 0.51 PL DWh + max.(HS_TOP, HS_BOT)+PI(20) 21.50 57.06 0.38 Pm ( or PL)+Pb+Q DWh + max.(HS_TOP, HS_BOT)+PI(20)

NA 11 Test Vertical Pm max. (PI(23)+155 kips, PE(14.7)+155 kips) 3.97 17.50 0.23 Pm+Pb max. (PI(23)+155 kips, PE(14.7)+155 kips) 9.00 26.25 0.34 PL max. (PI(23)+155 kips, PE(14.7)+155 kips) 3.21 26.25 0.12 Pm ( or PL)+Pb+Q max. (PI(23)+155 kips, PE(14.7)+155 kips)

NA 12 D

Horizontal External Pressure 21.70 45.10 0.48 13 D

Horizontal(2)

Pm PI(130) 11.83 44.3 0.27 Pm+Pb PI(130) 18.68 57.0 0.32 PL PI(130) 15.37 57.0 0.26 Notes:

(1)

DSC in transfer cask in vertical orientation. Only inner top cover is installed (2)

DSC in TC with TC in a horizontal orientation.

(3)

DSC in EOS-HSM supported on the steel rails All Indicated Changes are in response to Revised RAI 8-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 3.9.1-46 Table 3.9.1-7b DSC Shell Stress Results, Confinement Boundary - Load Combinations (Supported by FPS DSC Support Structure) 3 Pages Load Comb No.

Service Level DSC Orientation Stress Category Loads Stress intensity (ksi)

Allowable Stress (ksi)

Stress Ratio 9

A Horizontal(3)

Pm DWh + PI(20) 4.76 17.50 0.27 Pm+Pb DWh + PI(20) 10.20 26.25 0.39 PL DWh + PI(20) 14.02 26.25 0.53 Pm ( or PL)+Pb+Q DWh + PI(20) 21.11 52.50 0.40 Pm ( or PL)+Pb+Q+Pe DWh + PI(20) +TH 38.65 52.50 0.74 10 D

Horizontal(3)

Pm DWh + max.(HS_TOP, HS_BOT)+PI(20) 12.55 44.38 0.28 Pm+Pb DWh + max.(HS_TOP, HS_BOT)+PI(20) 32.29 57.06 0.57 PL DWh + max.(HS_TOP, HS_BOT)+PI(20) 23.99 57.06 0.42 Pm ( or PL)+Pb+Q DWh + max.(HS_TOP, HS_BOT)+PI(20)

NA 11 Test Vertical Pm max. (PI(23)+155 kips, PE(14.7)+155 kips) 3.97 17.50 0.23 Pm+Pb max. (PI(23)+155 kips, PE(14.7)+155 kips) 9.00 26.25 0.34 PL max. (PI(23)+155 kips, PE(14.7)+155 kips) 3.21 26.25 0.12 Pm ( or PL)+Pb+Q max. (PI(23)+155 kips, PE(14.7)+155 kips)

NA 12 D

Horizontal External Pressure 21.70 45.10 0.48 13 D

Horizontal(2)

Pm PI(130) 11.83 44.3 0.27 Pm+Pb PI(130) 18.68 57.0 0.32 PL PI(130) 15.37 57.0 0.26 Notes:

(1)

DSC in transfer cask in vertical orientation. Only inner top cover is installed.

(2)

DSC in transfer cask, TC is in horizontal orientation.

(3)

DSC in EOS-HSM-FPS supported by FPS DSC support structure.

(4)

Only load combinations 9 and 10 have results that differ from Table 3.9.1-7.

All Indicated Changes are in response to Revised RAI 8-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 6-17 For simplicity of input preparation in the TPA calculation, no credit is taken for down time between cycles (typically assumed to be 30 days). Using approximately 12 cycles to achieve a burnup of 300 GWd/MTU, the conservatism of this assumption is 11*30 = 330 days of uncredited decay time.

Results for Co-60 activity and decay heat for both the BPRA and TPA are summarized in Table 6-36 for a cooling time of 2 years. It is observed that the BPRA source may be used in the active fuel region, as the TPA does not extend into this region.

However, the TPA has a larger source than the BPRA in the plenum and top regions due to the high TPA burnup. Decay heat for both is small compared to SFA but must be accounted for during loading. The CC source used in the detailed PWR dose rate calculations is a hybrid CC source that combines the active fuel source of the BPRA with the top/plenum source of the TPA. This source is provided in Table 6-37. The CC source significantly impacts the peak dose rates on the side of the EOS-TC, due to the reduced lead thickness near the top nozzle.

TS Table 3 [6-11] is expressed in terms of Co-60 equivalent because some CC designs feature dose rate producing isotopes in addition to Co-60 (e.g., control rods containing silver or hafnium). For these CC types, it is convenient to convert the active fuel region CC sources to the equivalent activity of Co-60 that results in the same dose rate. This approach provides a common baseline for comparison against TS Table 3 [6-11]. For the top/plenum regions, Co-60 equivalent is simply the Co-60 activity because the top/plenum regions are always dominated by Co-60. Also, for CCs that contain primarily Co-60 in the active region (e.g., BPRAs), the Co-60 equivalent is simply the Co-60 activity.

The Co-60 equivalence methodology is generally applicable only for control rods containing silver or hafnium. The fuel assembly source terms developed in Section 6.2.2 are selected using simplified MCNP models (i.e., response functions) to estimate the dose rate on the side of the transfer cask or EOS-HSM roof vent (without vent covers). [

]

All Indicated Changes are a result of Enclosure 11, Item 2

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 6-18 6.2.5 Blended Low Enriched Uranium Fuel 6.2.6 Reconstituted Fuel All Indicated Changes are a result of Enclosure 11, Item 2

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 6-107 All Indicated Changes are a result of Enclosure 11, Item 2 Proprietary Information on Pages 6-107 through 6-108 Withheld Pursuant to 10 CFR 2.390

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 8-24 Outokumpu, Data Sheet Duplex Stainless Steel.

8-48 Euronorm EN485-2, Aluminium and aluminium alloys - Sheet, strip and plate - Part 8-49 2: Mechanical properties.

Euronorm EN573-3, Aluminium and aluminium alloys - Chemical composition and 8-50 form of wrought products - Part 3: Chemical composition and form of products.

The Steel Construction Institute, Design Manual for Structural Stainless Steel, 4th 8-51 Edition, 2017.

All Indicated Changes are in response to Revised RAI 8-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 8-31 Table 8-7 Material Properties, SA-240/SA-479 Type 2205 / SA-182 Gr F60 (1)

Temp

(°F)

E (103 ksi)

Sm (ksi)

Sy (ksi)

Su (ksi)

AVG (10-6 °F -1)

(lb/in3)

K (Btu/hr-ft-°F)

Cp (Btu/lb-°F)

-20 37.5 65.0 90.0 0.280 70 29.0 37.5 65.0 90.0 7.0 8.2 0.122 100 37.5 65.0 90.0 7.1 8.3 0.122 150 37.5 60.5 7.2 8.6 0.123 200 28.2 37.5 57.8 90.0 7.3 8.8 0.125 250 37.0 55.5 7.3 9.1 0.128 300 27.5 35.8 53.7 86.8 7.4 9.3 0.128 350 34.9 7.5 9.5 0.129 400 27.0 34.2 51.2 83.5 7.6 9.8 0.131 450 7.6 10.0 0.132 500 26.4 49.6 81.6 7.7 10.2 0.132 550 7.8 10.5 0.134 600 26.0 47.9 80.7 7.8 10.7 0.134 650 46.9 80.5 7.9 10.9 0.135 700 25.5 7.9 11.2 0.136 752 39.0 (2) 73.8 (2) 8.0 11.4 0.137 800 25.1 8.0 11.6 0.137 850 8.1 11.9 0.139 900 8.1 12.1 0.139 950 8.2 12.3 0.140 1000 8.2 12.5 0.140 ASME Table TM-1

p. 738 Group H Table 5A

p. 420, Lines 29, 30 Table Y-1

p. 632 Lines 1, 2, 3

Table U, pp. 502-503, Lines 7, 8, 9 Table TE-1

p. 708 Group 2 Ref. 8-48 Calculated based on Table TCD

p. 728 Group K Source: ASME Section II, Part D Note:

1.

These properties are provided for information only; the design analysis is based on the mechanical properties and thermal conductivity of SA-240 Types 304 and 316, which bound the duplex steel

2.

Strength properties greater than 650°F are based on the rate of reduction from the Design Manual for Structural Stainless Steel [8-51]

All Indicated Changes are in response to Revised RAI 8-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 8-32 Table 8-8 Material Properties, SA-240 UNS S31803 / SA-182 Gr F51 (1)

Temp

(°F)

E (103 ksi)

Sm (ksi)

Sy (ksi)

Su (ksi)

AVG (10-6 °F -1)

(lb/in3)

K (Btu/hr-ft-°F)

Cp (Btu/lb-°F)

-20 90 0.280 70 29.0 90 7.0 8.2 0.122 100 30 65 90 7.1 8.3 0.123 150 7.2 8.6 0.125 200 28.2 30 57.8 90 7.3 8.8 0.125 250 7.3 9.1 0.128 300 27.5 28.9 53.7 86.8 7.4 9.3 0.128 350 7.5 9.5 0.129 400 27.0 27.8 51.2 83.5 7.6 9.8 0.131 450 7.6 10.0 0.132 500 26.4 27.2 49.6 81.6 7.7 10.2 0.132 550 7.8 10.5 0.134 600 26.0 26.9 47.9 80.7 7.8 10.7 0.134 650 7.9 10.9 0.135 700 25.5 7.9 11.2 0.136 752 39.0 (2) 73.8 (2) 8.0 11.4 0.137 800 25.1 8.0 11.6 0.137 850 8.1 11.9 0.139 900 8.1 12.1 0.139 950 8.2 12.3 0.140 1000 8.2 12.5 0.140 ASME Table TM-1

p. 738 Group H ASME Code Case N-635-1 ASME Code Case N-635-1 Table U Table TE-1

p. 708 Group 2 Ref. 8-48 Calculated based on Table TCD

p. 728 Group K Sources: ASME Section II, Part D and ASME Code Case N-635-1 Note:

1.

These properties are provided for information only; the design analysis is based on the mechanical properties and thermal conductivity of SA-240 Types 304 and 316, which bound the duplex steels.

2.

Strength properties greater than 600°F are based on the rate of reduction from the Design Manual for Structural Stainless Steel [8-51].

All Indicated Changes are in response to Revised RAI 8-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 12-7 Components UFSAR Sections EOS-TC loaded with EOS-37PTH DSC (Side and corner drop)

Appendices 3.9.1 -

3.9.3 and 3.9.5 EOS-TC loaded with EOS-89BTH DSC (Side and corner drop)

Appendices 3.9.1 -

3.9.3 and 3.9.5 EOS-37PTH DSC, PWR Fuel Cladding (Side and corner drop)

Appendix 3.9.6 EOS-89BTH DSC, BWR Fuel Cladding (Side and corner drop)

Appendix 3.9.6 All stresses are within allowable limits in both drop scenarios for the TC108, EOS-37PTH DSC and EOS-89BTH DSC. The largest strain in the basket is 0.9 %

and 0.6 % for the EOS-TC108 loaded with EOS-37PTH DSC and EOS-89BTH DSC, respectively. The maximum stresses and strains in the fuel cladding for the side and corner drops remain below the applicable yield strength, therefore there is no fuel deformations.

The strain is limited in effect given the mode of deformation as the basket plates maintain their general shape. This deformation is limited and the position of the fuel assemblies is maintained from their initial positions relative to each other. The deformations of the basket plates are approximately uniform in the direction of impact and the fuel does not change configuration. Therefore, these deformations do not have an effect on criticality control.

Accident Dose Calculation Based on analysis results presented in Appendix 3.9.3, Sections 3.9.3.3 and 3.9.3.4, the accidental EOS-TC drop scenarios do not breach the EOS-37PTH or the EOS-89BTH DSC confinement boundaries. The function of EOS-TC lead shielding is not compromised by these drops. The EOS-TC neutron shield, however, may be damaged in an accidental drop.

Dose rates are computed at 100 m from the EOS-TC with the neutron shield removed, which is the minimum allowed distance to the site boundary. As presented in Chapter 6, Table 6-54, the maximum dose rate at 100 m from an EOS-TC during a loss of neutron shield and lead slump in an accidental drop accident is 2.89 mrem/hr. Based on the discussion in Section 6.2.8, the maximum accident dose rate is doubled to 5.8 mrem/hr. If an 8-hour recovery time is assumed, the dose to an individual at the site boundary is 5.8*8 = 46 mrem, which is significantly below the 10 CFR 72.106 dose limit of 5 rem.

Corrective Actions Recovery actions for an accident condition where the neutron shield and air circulation are lost would include the following actions immediately after an accident:

All Indicated Changes are in response to Revised RAI 8-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 12-8

1. Re-instate neutron shielding to protect workers on the site and/or restart air circulation to control the fuel cladding and the DSC temperatures if necessary based on the heat load of the DSC.

a) If the neutron shield panels are punctured or damaged to the point that the neutron shielding cannot be refilled, appropriate supplementary shielding should be utilized, considering air flow, TC/DSC temperatures, and type of shielding required. This may include lead blankets, concrete barriers, metal sheeting, or other neutron shielding mechanisms.

b) If the initial blower was damaged in the accident, the air circulation system is equipped with a redundant blower to facilitate engagement of air circulation.

Caution: Air circulation or neutron shielding should be reinstated as soon as possible, if necessary based on the heat load of the DSC.

2. Evaluate the TC and DSC condition to determine if they are safe to move. This evaluation should include a transient thermal evaluation to determine the time the air circulation and neutron shield water, or both, should be in place to cool the DSC to a safe condition for transfer. The evaluation would consider the accident specifics such as the decay heat load, the ambient temperature, and the time from loss of cooling or neutron shield water to the time the cooling or water are re-introduced, etc. This will reduce the temperatures and provides additional margins to allow for safe transfer.

3. Return the TC/DSC to the fuel building or other acceptable location for evaluation of continued service after evaluations have determined that the TC is safe to move. Upright the TC and fill the annulus with water. Monitor the annulus water as required to ensure continued DSC cooling until a further determination regarding continued serviceability of the TC, DSC, and contents is reached.

4. The DSC is then opened and the fuel removed for inspection, as necessary.

Removal of the EOS-TC top cover plate may require cutting of the bolts in the event of a corner drop onto the top end. These operations take place in the plant fuel building decontamination area and spent fuel pool after recovery of the EOS-TC.

5. Following recovery of the EOS-TC and unloading of the DSC, the EOS-TC is inspected, repaired and tested as appropriate prior to reuse. For recovery of the EOS-TC and contents, it may be necessary to develop a special sling/lifting apparatus to move the EOS-TC from the drop site to the fuel pool. This may require several weeks of planning to ensure all steps are correctly organized.

During this time, lead blankets may be added to the EOS-TC to minimize onsite exposure to site operations personnel. The EOS-TC can be roped off to ensure the safety of the site personnel.

All Indicated Changes are in response to Revised RAI 8-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 12-12 The factor of safety on tip is 1.30 from the bounding DBT wind plus missile load combination on the EOS-TC while sitting on the trailer ready for transfer. The primary membrane intensity and combined membrane plus bending stresses due to DBT and missile impact are calculated, which are below the allowable stresses. The maximum missile penetration depth is found to be 0.526 inch, which is less than the thickness of EOS-TC outer shell of 1.0 inch and top cover plate thickness of 3.25 inches.

Accident Dose Calculation Based on the above analyses, the DSC confinement boundary is not breached as a result of the missile impacts. Accordingly, no DSC damage or release of radioactivity is postulated.

The missile impact scenario may result in the loss of EOS-TC neutron shielding and local deformation/damage of the gamma shielding. The effect of loss of the neutron shielding due to a missile impact is bounded by that resulting from a EOS-TC drop scenario evaluated above in Section 12.3.1. The change in radiation dose due to local deformation/damage of the gamma shielding is negligible.

Corrective Action After excessive high winds or a tornado, the EOS-TC is inspected for damage. These operations take place in the plant fuel building decontamination area and spent fuel pool after recovery of the EOS-TC. If the neutron shield or the air circulation on the EOS-TC was lost, and time limits for transfer exceeded, follow the corrective actions detailed in Section 12.3.1 for the accident drop. The transfer cask and DSC would be moved only after these evaluations determined it was safe to do so. Following recovery of the EOS-TC and unloading of the DSC, the EOS-TC is inspected, repaired and tested as appropriate prior to reuse.

For recovery of the EOS-TC and contents, it may be necessary to develop a special sling/lifting apparatus to move the EOS-TC from the site to the fuel pool. This may require several weeks of planning to ensure all steps are correctly organized. During this time, lead blankets or other appropriate supplementary shielding may be added to the EOS-TC to minimize on-site exposure to site operations personnel. The EOS-TC can be roped-off to ensure the safety of the site personnel.

12.3.5 Flood Cause of Accident Flooding conditions simulating a range of flood types, such as tsunami and seiches as specified in 10 CFR 72.122 (b) are considered. In addition, floods resulting from other sources, such as high water from a river or a broken dam, are postulated as the cause of the accident.

All Indicated Changes are in response to Revised RAI 8-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. 3, 06/20 October 2020 Revision 6 72-1042 Amendment 2 Page 12-16 Corrective Actions Evaluation of EOS-TC neutron shield damage as a result of a fire is to be performed to assess the need for temporary shielding (if fire occurs during transfer operations) and repairs to restore the EOS-TC to pre-fire design conditions.

Similarly, the DSC should be inspected for damage if the EOS-TC neutron shield were damaged or the air circulation on the EOS-TC was lost and time limits for transfer exceeded. For these accident conditions, follow the corrective actions detailed in Section 12.3.1 for the accident drop.

All Indicated Changes are in response to Revised RAI 8-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. TBD, TBD October 2020 Revision 6 72-1042 Amendment 2 Page B.4-1 Appendix B is newly added in Amendment 2.

B.4 THERMAL EVALUATION The thermal evaluation described in this chapter is to demonstrate that a 61BTH Type 2 Dry Shielded Canister (DSC) can be loaded inside the OS197 transfer cask (TC) and the NUHOMS MATRIX (HSM-MX) for normal, off-normal, and accident conditions while maintaining temperatures and pressures within the specified regulatory limits per NUREG-1536 [B.4-1].

A summary of the 61BTH Type 2 DSC configuration analyzed in this chapter is shown below:

DSC Type Basket Assembly Type or Heat Load Zone Configuration (HLZC)

Max.

Heat Load (kW)

Neutron Absorber Material Transfer Cask Time Limit for Transfer Operation Storage Module 61BTH Type 2 1, 2, 9 22.0 Borated Aluminum or BORAL or MMC OS197 or OS197H or OS197FC-B No NUHOMS MATRIX (HSM-MX) 3, 4 19.4 8

27.4 OS197FC-B Yes 5, 6, 7, 10 31.2 Borated Aluminum or MMC NOTE: The 61BTH Type 2 DSC and the OS197 TCs (OS197, OS197H, or OS197FC-B) are licensed under Revision 18 of the UFSAR of CoC 1004 in Appendix T [B.4-2] and no changes are considered.

As shown in the above table, the 61BTH Type 2 DSC will be transferred in an OS197 TC and stored in an HSM-MX. Since the combination of the 61BTH Type 2 DSC and the OS197 TC for transfer operations is approved for use in the CoC 1004 and there are no changes to the components, the thermal analyses performed for transfer operations presented in Appendix T.4 of the CoC 1004 UFSAR for the NUHOMS general license [B.4-2] are applicable. In addition to these evaluations, Section B.4.5.6 presents new evaluations to determine the duration required to run the air circulation and the duration available to complete the transfer operations once the air circulation is turned off. For storage operations of the 61BTH Type 2 DSC in the HSM-MX, new analyses are presented in Section B.4.4 since this combination was not previously analyzed.

The 61BTH Type 2 DSC is analyzed based on a maximum heat load of 31.2 kW from 61 boiling water reactor (BWR) fuel assemblies (FAs) with a maximum heat load of 1.2 kW per assembly. A total of ten HLZCs shown in Figures 4A through 4J of the Technical Specification [B.4-3] are authorized in the 61BTH Type 2 DSC. The location of damaged and failed FAs inside the 61BTH Type 2 DSC is also provided in Figure 5 of the Technical Specification [B.4-3]. HLZCs considered for the 61BTH Type 2 DSC including the placement of damaged/failed fuel assemblies are identical to those previously evaluated in Appendix T, Section T.4 of [B.4-2].

All Indicated Changes are in response to Revised RAI 4-7

NUHOMS EOS System Updated Final Safety Analysis Report Rev. TBD, TBD October 2020 Revision 6 72-1042 Amendment 2 Page B.4-14 Appendix B is newly added in Amendment 2.

B.4.3 Thermal Loads and Environmental Conditions For storage operations in the HSM-MX, the normal ambient temperature is considered in the range of -20 °F to 100 °F. A daily average ambient temperature of 90 °F is used in the evaluations, corresponding to a daily maximum temperature of 100 °F for the normal hot storage conditions as discussed in Chapter A.4, Section A.4.3.

Off-normal ambient temperature is considered in the range of - 40 °F to 117 °F. A daily average ambient temperature of 103 °F is used in the evaluations, corresponding to a daily maximum temperature of 117 °F for the off-normal hot storage conditions as discussed in Chapter A.4, Section A.4.3.

Ambient temperatures of -20 F and -40 F are considered for the normal and off-normal cold storage conditions, respectively.

The HSM-MX is located outdoors and is exposed to the environment. Wind is a normal environment variable that varies frequently both in direction and magnitude.

For the HSM-MX, low speed wind in the range of 0 to 15 mph is considered for normal storage conditions based on the discussion in Section 2.5 of NUREG-2174

[B.4-8].

Summary of OS197FC-B load cases for the 61BTH Type 2 DSC are provided in Appendix T, Table T.4-4, and Table T.4-5 of [B.4-2] and Section B.4.5.6.1. The ambient temperature ranges are 0 to 100 °F (-17.8 to 37.8 °C) for normal transfer and 0 to 117 °F (-17.8 to 47.2 °C) for off-normal transfer operations. The indoor ambient temperature of 120 °F is considered for fuel loading operations in the fuel building.

All Indicated Changes are in response to Revised RAI 4-7

NUHOMS EOS System Updated Final Safety Analysis Report Rev. TBD, TBD October 2020 Revision 6 72-1042 Amendment 2 Page B.4-21 Appendix B is newly added in Amendment 2.

B.4.5 Thermal Evaluation for Transfer Casks with 61BTH Type 2 DSC The OS197 TCs (OS197H, OS197FC, and OS197FC-B) are used to transfer the loaded 61BTH Type 2 DSC between the fuel building and the ISFSI. Thermal evaluation of the 61BTH Type 2 DSCs in the OS197 TCs is presented in Appendix T, Chapter T.4 of [B.4-2]. These evaluations are applicable for transfer operations due to the following reasons:

1. 61BTH Type 2 DSC and OS197 TC are identical to the design previously evaluated in Appendix T of CoC 1004 [B.4-2] as discussed in Chapter B.1, Section B.1.1.

2. The combination of DSC/TC for transfer operations remains unchanged from those evaluated in Appendix T, Chapter T.4 of [B.4-2].

3. HLZCs considered for the 61BTH Type 2 DSC including the placement of damaged/failed fuel assemblies are identical to those previously evaluated in Appendix T, Chapter T.4 of [B.4-2].

4. Ambient conditions for transfer operations presented in Section B.4.3 for transfer operations remain identical to the ambient conditions considered in Appendix T, Section T.4.5.2 of [B.4-2].

5. Various load cases evaluated for transfer operations in Appendix T, Section T.4.5.2 of [B.4-2] remain applicable without any changes.

6. Time limits for transfer operations if necessary before initiation of a recovery action such as air circulation are evaluated in Appendix T, Section T.4.5.4 of

[B.4-2].

If air circulation is chosen as a recovery option, additional time limits that specify the minimum duration required to run the air circulation and also the maximum duration available to complete transfer operations once the air circulation is turned off are determined in Section B.4.5.6.

Background of Transfer Evaluation in CoC 1004 Thermal evaluation for transfer operations for the 61BTH Type 2 DSCs in the OS197 TCs in Appendix T, Chapter T.4 of [B.4-2] are performed using a two-step approach.

The two steps are:

1. Thermal evaluation of the OS197TCs with the 61BTH DSC is performed to determine the DSC shell temperature profile and the maximum component temperatures of the OS197 TC components.

In this step, the fuel basket and the hold down ring within the 61BTH Type 2 DSC are modeled as homogenous solids. Since the fuel basket is modeled as homogenous solids, this step only considers the total heat load per DSC and does not depend on the individual HLZCs. Based on a review of the various HLZCs, two sets of evaluations are performed. The first set is for transfer operations with heat loads 22.0 kW and the second set is for transfer operations with heat loads

> 22.0 kW and 31.2 kW.

All Indicated Changes are in response to Revised RAI 4-7

NUHOMS EOS System Updated Final Safety Analysis Report Rev. TBD, TBD October 2020 Revision 6 72-1042 Amendment 2 Page B.4-23 Appendix B is newly added in Amendment 2.

The OS197FC-B TC contains design provisions for the use of air circulation system to improve its thermal performance for heat loads greater than 22.0 kW for 61BTH Type 2 DSC. The air circulation system consists of redundant, industrial grade pressure blowers and power systems, ducting, etc. When operating, the fan system is expected to generate a flow rate of 400 cfm or greater, which will be ducted to the location of the ram access cover at the bottom of the TC. The air circulation system is not needed for heat loads 22.0 kW. Section B.4.5.6 establishes the minimum duration required to operate the air circulation. It also evaluates the duration available once the air circulation is turned off to transfer the DSC to the storage module. This evaluation is based on 61BTH Type 2 DSC with maximum allowable heat load of 31.2 kW. If the maximum heat load of a DSC is less than 31.2 kW, new time limits may be determined to provide additional time for these transfer operations.

Section B.4.5.1.1 presents a discussion on the various load cases considered in the thermal evaluation of the 61BTH Type 2 DSC during transfer operations in the OS197 TCs.

Section B.4.5.1.2 presents a description of the model used for the thermal evaluation of the 61BTH Type 2 DSC during the transfer in the OS197 TCs.

Section B.4.5.1.3 presents the results of the thermal evaluation for normal, off-normal, and hypothetical accident conditions of transfer for the OS197 TCs with heat loads

>22.0 kW and 31.2 kW in the 61BTH Type 2 DSC.

B.4.5.1.1 61BTH Type 2 DSC - Description of Load Cases for Transfer Various load cases are considered to determine the thermal performance of the OS197 TCs with the 61BTH Type 2 DSC described in Appendix T, Section T.4.5.2 of

[B.4-2]. The load cases are further listed in the following tables:

1. Table T.4-4 of [B.4-2] for a maximum decay heat load of 22.0 kW,

2. Table T.4-5 of [B.4-2] for a maximum decay heat load of 31.2 kW.

The load cases considered for transfer of the 61BTH Type 2 DSC include the vertical loading condition inside of the fuel handling facility, normal and off-normal horizontal transfer conditions with and without air circulation outside the fuel handling facility, and hypothetical accident scenarios.

It should be noted that the thermal evaluations in Appendix T, Chapter T.4 of [B.4-2]

for heat loads 22.0 kW are based on the 61BTH Type 1 DSC. The 61BTH Type 2 DSC uses aluminum R90 rail in place of the steel plate rail in the 61BTH Type 1 DSC and is thermally more efficient than the 61BTH Type 1 DSC. Therefore, the thermal evaluation for the OS197 TCs with the 61BTH Type 1 DSC reported in Appendix T, Section T.4.5 of [B.4-2] represents the bounding thermal evaluation for the OS197 TCs with heat loads 22.0 kW in the 61BTH Type 2 DSC.

All Indicated Changes are in response to Revised RAI 4-7

NUHOMS EOS System Updated Final Safety Analysis Report Rev. TBD, TBD October 2020 Revision 6 72-1042 Amendment 2 Page B.4-24 Appendix B is newly added in Amendment 2.

For the five HLZCs (HLZC 1 through HLZC 4 and HLZC 9) with heat loads 22.0 kW allowed for the 61BTH Type 2 DSCs as shown in Figure 4A through Figure 4D and Figure 4I of the Technical Specifications [B.4-3], steady-state transfer operations are permitted.

For the five HLZCs (HLZC 5 through HLZC 8 and HLZC 10) with heat loads >22.0 kW and 31.2 kW allowed for the 61BTH Type 2 DSCs as shown in Figure 4E through Figure 4H and Figure 4J of the Technical Specifications [B.4-3], time limits are established to complete the normal and off-normal transfer operations to ensure that the temperature limits for the various components described in Section B.4.2 are not exceeded. There are no time limits associated with accident conditions that are evaluated at steady-state.

If the transfer operations for the five HLZCs (HLZC 5 through HLZC 8 and HLZC 10) cannot be completed within the time limits established in Technical Specifications

[B.4-3], one of the recovery options is to initiate the air circulation. Section B.4.5.6.1 presents additional load cases that are evaluated if air circulation is initiated.

Table B.4-9 presents the load cases for these evaluations.

For all the normal, off-normal hot conditions, and accident design load cases considered in Tables T.4-4 and T.4-5 of [B.4-2], insolation is considered per 10 CFR 71.71 [B.4-9].

B.4.5.1.2 Thermal Model of OS197FC-B TC There is no change to the thermal model of the OS197FC-B TC with the 61BTH Type 2 DSC described in Appendix T, Section T.4.5.1 of [B.4-2].

The SINDA/FLUINT' [B.4-10] and Thermal Desktop [B.4-11] computer codes described in Appendix T, Section T.4.5.1.1 of [B.4-2] are used to model the OS197FC-B TC (or OS197/OS197H TC) with the 61BTH DSCs to determine the temperature distribution in the TC and the DSC shell.

If air circulation is initiated as one of the recovery options, the thermal model described in Section B.4.5.6.2 is used to evaluate the thermal performance.

B.4.5.1.3 OS197 TC Thermal Model Results The maximum temperature results for the 61BTH DSC shell assemblies and TC components during transfer are discussed in Appendix T.4, Section T.4.5.3 and presented in Table T.4-7 through Table T.4-9 of Appendix T.4 of [B.4-2]. These results are for 31.2 kW and 22.0 kW heat loads, respectively. The DSC shell temperatures are used as boundary conditions in the 61BTH DSC thermal analysis presented in Section B.4.5.2 to calculate the basket and fuel cladding temperatures.

All Indicated Changes are in response to Revised RAI 4-7

NUHOMS EOS System Updated Final Safety Analysis Report Rev. TBD, TBD October 2020 Revision 6 72-1042 Amendment 2 Page B.4-25 Appendix B is newly added in Amendment 2.

B.4.5.1.3.1 Normal and Off-Normal Transfer without Forced Air Circulation (FC)

There is no change to the normal and off-normal transfer evaluations described in Appendix T.4, Section T.4.5.3.1 of [B.4-2] without FC.

Steady-State analyses are performed to determine the maximum temperature results listed in Table T.4-7 of Appendix T.4 of [B.4-2] for the 61BTH DSC shell assemblies and TC components for DSC transfer under normal and off-normal operations with a decay heat load 22.0 kW.

Transient analyses are performed to determine the time limit for DSC transfer operations for 61BTH Type 2 DSCs with a decay heat load higher than 22.0 kW up to 31.2 kW. The transient maximum temperature results of the 61BTH DSC shell assemblies and TC components for DSC transfer without FC under normal and off-normal operations with a decay heat load above 22.0 kW are listed in Appendix T.4, Table T.4-8 of [B.4-2].

Based on targeted DSC shell temperatures of approximately 405 °F (for HLZCs 7 and

10) and 445 ºF (for HLZCs 5, 6, and 8) to avoid excessive fuel cladding temperatures, the transient analysis indicates that time limits approximately 15 and 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />, respectively, are available to transfer the DSC into the HSM-MX or take some other corrective actions. The anticipated corrective actions are described in Appendix T.4, Section T.4.5.3.1 of [B.4-2].

The results from Section B.4.5.2 documented in Tables T.4-12 and T.4-17 of [B.4-2]

show that, even with these shell temperatures for normal and off-normal transfer conditions, there is considerable margin in the bounding cladding temperatures (734 ºF and 722 ºF calculated for normal and off-normal cases, respectively, vs. a 752 ºF limit).

B.4.5.1.3.2 Normal and Off-Normal Transfer with Forced Air Circulation The normal and off-normal transfer evaluations described in Appendix T.4, Section T.4.5.3.2 of [B.4-2] are applicable under steady-state conditions when air circulation is enabled.

For the transfer time periods exceeding the specific time limits above 22.0 kW, one of the corrective actions available to limit the temperature increase is to initiate air circulation in the TC/DSC annulus.

Table T.4-9 of Appendix T.4 of [B.4-2] presents the maximum component temperatures achieved under bounding normal and off-normal ambient operating conditions for the OS197FC-B TC with a 61BTH DSC with 31.2 kW of decay heat and a flow rate of 400 cfm of air circulation. As seen, all component temperatures are below their limits.

All Indicated Changes are in response to Revised RAI 4-7

NUHOMS EOS System Updated Final Safety Analysis Report Rev. TBD, TBD October 2020 Revision 6 72-1042 Amendment 2 Page B.4-26 Appendix B is newly added in Amendment 2.

Section B.4.5.6 presents additional analyses that determine the minimum duration required to run the air circulation and also the maximum duration available to complete transfer operations once air circulation is turned off. Table B.4-10 presents the results for these evaluations.

B.4.5.1.3.3 Accident Transfer There is no change to the accident transfer thermal results presented in Appendix T.4, Section T.4.5.3.3 of [B.4-2].

Based on the discussion in Appendix T.4, Section T.4.5.3.3 of [B.4-2], loss of neutron shield is the bounding accident condition. Table T.4-10 of [B.4-2] presents the peak component temperatures achieved under this accident at steady-state conditions.

B.4.5.1.4 Evaluation of OS197FC-B TC Performance There is no change to the evaluation presented in Appendix T, Section T.4.5.4 of

[B.4-2] on the thermal performance of the OS197FC-B TC for normal, off-normal, and accident conditions of operation when heat loads are less than or equal to 22 kW.

For heat loads > 22kW and 31.2 kW, the transfer time limits of 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> and 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> specified in Appendix T.4.5.4 of [B.4-2] are based on a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> recovery time.

However, to be consistent with the EOS-37PTH and EOS-89BTH DSCs the recovery time to complete the various action statements in LCO 3.1.3 of the Technical Specifications [B.4-3] is increased by 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> to a total of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> with corresponding reduction in the transfer time limits. Therefore, the time limits for EOS-61BTH DSC are reduced to 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> and 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> based on the HLZC.

Based on the discussion in Section 4.5.4, if air circulation cannot be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after exceeding the transfer time limit, the TC/DSC has to be returned to the cask handling area to be positioned in vertical orientation and then the TC/DSC annulus will be filled with clean water. As discussed in Section 4.5.4, a total of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is available to complete Action A.2 and Action A.3 of the LCO 3.1.3 of the Technical Specifications [B.4-3] with a maximum duration of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for Action A.2.

The allowable duration for the transfer operations (defined as from the time when the water in the TC-DSC annulus is drained to when the DSC is loaded into the storage module) will vary depending only on the DSC type and the heat load configuration.

For simplicity of operations, a single time limit is used for all ambient conditions and TC orientations (i.e., longer times are available for the non-controlling conditions).

The following table summarizes the permissible operational conditions:

DSC Heat Load Zoning Configuration Transfer Time Limit (1), (2) (4)

HLZCs 1, 2,3, 4 and 9 (5) ( 22 kW)

No time limit HLZCs 5, 6 ( 31.2 kW) 23.0 Hours (3)

HLZCs 7, 10 (5) ( 31.2 kW) 10.0 Hours (3)

HLZC 8 ( 27.4 kW) 23.0 Hours (3)

Notes:

All Indicated Changes are in response to Revised RAI 4-7

NUHOMS EOS System Updated Final Safety Analysis Report Rev. TBD, TBD October 2020 Revision 6 72-1042 Amendment 2 Page B.4-27 Appendix B is newly added in Amendment 2.

(1) Transfer time is defined as from the time when the TC/DSC annulus water is drained to when the DSC is loaded into the storage module.

(2) The listed allowable transfer times are valid for all ambient conditions and TC orientations.

(3) Initiate recovery operations such as air circulation if the operation time exceeds the limit per LCO 3.1.3 of Technical Specifications [B.4-3].

(4) The transfer operation time limit is reset only if the transfer cask annulus is refilled with water.

(5) Thermal evaluation of 61BTH DSC for HLZCs 9 and 10 is presented in Section T.4.6.10 of [B.4-2].

B.4.5.2 61BTH DSC Thermal Analysis Thermal analysis of the 61BTH DSC for transfer operations is described in Appendix T, Section T.4.6 of [B.4-2]. In addition to these evaluations, Section B.4.5.6 presents additional evaluations to determine the maximum fuel cladding temperatures and basket component temperatures if air circulation is initiated.

B.4.5.2.1 Heat Load Zoning Configurations There is no change to the HLZCs allowed within the 61BTH Type 2 DSC. A total of 10 HLZCs are allowed for the 61BTH DSCs as shown in Figure 4A through Figure 4J of the Technical Specification [B.4-3]. Thermal evaluation of the 61BTH Type 2 DSC with HLZCs 1 through 8 are presented in Appendix T, Section T.4.6.1 through Section T.4.6.9 of [B.4-2].

Thermal evaluation of the 61BTH Type 2 DSC for HLZCs 9 and 10 is presented in Appendix T, Section T.4.6.10 of [B.4-2].

It should be noted that the thermal evaluations in Appendix T.4 of [B.4-2] for heat loads 22.0 kW are based on the 61BTH Type 1 DSC. The 61BTH Type 2 DSC uses aluminum R90 rail in place of the steel plate rail in the 61BTH Type 1 DSC and is thermally more efficient than the 61BTH Type 1 DSC. Therefore, the thermal evaluation for the OS197 TCs with the 61BTH Type 1 DSC reported in Appendix T, Section T.4.5 of [B.4-2] represents the bounding thermal evaluation for the OS197 TCs with heat loads 22.0 kW in the 61BTH Type 2 DSC.

B.4.5.2.2 61BTH DSC Thermal Model There is no change to the 61BTH Type 2 DSC thermal model described in Appendix T, Section T.4.6.2, T.4.6.3, T.4.6.4 and T.4.6.5 of [B.4-2].

If air circulation is initiated as one of the recovery options, the thermal model described in Section B.4.5.6.2 is used to evaluate the thermal performance.

B.4.5.2.3 61BTH Type 2 DSC Thermal Evaluation (HLZCs 1 through 8, Intact Fuel)

There is no change to the thermal evaluation of 61BTH Type 2 DSC transfer in the OS197 TCs described in Appendix T.4, Sections T.4.6.6, T.4.6.7 and T.4.6.8 of

[B.4-2] for normal, off-normal and accident conditions, respectively.

All Indicated Changes are in response to Revised RAI 4-7

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Section B.4.5.6 presents additional analyses that determine the minimum duration required to run the air circulation and also the maximum duration available to complete transfer operations once air circulation is turned off. Table B.4-10 presents the results for these evaluations.

Normal Transfer Evaluation The bounding maximum fuel cladding temperatures during normal transfer conditions are listed in Table T.4-12 of [B.4-2]. The maximum fuel cladding temperatures are well below the allowable fuel cladding temperature limit of 752 °F (400 °C) [B.4-1]

for short-term transfer operations.

The maximum temperatures of the basket assembly components for normal transfer conditions for the bounding HLZCs are listed in Tables T.4-13 and T.4-14 of [B.4-2]

for maximum heat loads per DSC up to 22.0 kW and 31.2 kW, respectively.

The DSC temperature distributions for normal transfer operations are shown in Figures T.4-29 and T.4-30 of [B.4-2] for 22.0 kW heat load and Figures T.4-33 and T.4-34 of [B.4-2] for 31.2 kW heat load, respectively.

Off-Normal Transfer Evaluation The bounding maximum fuel cladding temperatures during normal transfer conditions are listed in Table T.4-17 of [B.4-2]. The maximum fuel cladding temperatures are well below the allowable fuel cladding temperature limit of 752 °F (400 °C) [B.4-1]

for short-term transfer operations.

The maximum temperatures of the basket assembly components for off-normal transfer conditions for the bounding HLZCs are listed in Tables T.4-18 and T.4-19 of

[B.4-2] for maximum heat loads per DSC up to 22.0 kW and 31.2 kW, respectively.

The DSC temperature distributions for off-normal transfer operations are shown in Figures T.4-29 and T.4-30 of [B.4-2] for 22.0 kW heat load and Figures T.4-33 and T.4-34 of [B.4-2] for 31.2 kW heat load, respectively.

Accident Transfer Evaluation The maximum fuel cladding temperatures during accident transfer conditions are evaluated for all decay HLZCs as listed in Table T.4-21 of [B.4-2]. The maximum fuel cladding temperatures are well below the allowable fuel temperature limit of 1058 °F (570 °C) [B.4-1] for accident transfer operations.

The maximum temperatures of the basket assembly components for normal transfer conditions for the bounding HLZCs are listed in Tables T.4-22 and T.4-23 of [B.4-2]

for maximum heat load per DSC up to 22.0 kW and 31.2 kW heat load, respectively.

Figure T.4-31 of [B.4-2] shows the DSC temperature distributions for accident transfer operations with 22.0 kW heat load.

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NUHOMS EOS System Updated Final Safety Analysis Report Rev. TBD, TBD October 2020 Revision 6 72-1042 Amendment 2 Page B.4-30 Appendix B is newly added in Amendment 2.

Figure 5 of the Technical Specification [B.4-3] shows that the 61BTH DSC allows for the storage of up to 61 damaged fuel assemblies. For the worst case with 60 damaged FAs and one intact FA, the maximum fuel cladding temperature for intact FA is 955 °F. However, for all evaluations with intact FAs, the maximum fuel cladding temperatures are well below the limit of 1058 °F. For the case with 61 damaged fuel assemblies, since all damaged fuel assemblies are considered as rubble, there are no thermal limits associated with this scenario. Therefore, there is no impact on loading damaged fuel along with intact fuel within the 61BTH Type 2 DSC.

B.4.5.5 Thermal Evaluation for Loading/Unloading Conditions There is no change to the thermal evaluation for loading and unloading conditions presented in Appendix T, Section T.4.7 of [B.4-2].

B.4.5.5.1 Maximum Fuel Cladding Temperature during Vacuum Drying There is no change to the thermal evaluation during vacuum drying operations presented in Appendix T, Section T.4.7.1 of [B.4-2].

Tables T.4-25 and T.4-27 of [B.4-2] provide the maximum calculated temperatures for the fuel cladding and the basket components for the 61BTH Type 2 DSC during vacuum drying.

The maximum cladding temperatures for vacuum drying using helium are 598 °F for the 61BTH Type 2 DSC. This maximum cladding temperature is well below the limit of 752 °F [B.4-2].

B.4.5.5.2 Evaluation of Thermal Cycling of Fuel Cladding during Vacuum Drying, Helium Backfilling and Transfer There is no change to the discussion on thermal cycling of fuel cladding during vacuum drying operations presented in Section T.4.7.2 of [B.4-2].

B.4.5.5.3 Reflooding Evaluation There is no change to the discussion on unloading operations presented in Section T.4.7.3 of [B.4-2].

B.4.5.6 Minimum Duration to Operate Forced Air Circulation (FC)

Section B.4.5.1.4 summarizes the transfer time limits for OS197FC-B TC loaded with 61BTH Type 2 DSC with heat loads > 22.0 kW and 31.2 kW. If the transfer time limit cannot be satisfied, one of the recovery actions is to initiate Forced Air Circulation (FC). This section provides a thermal evaluation to establish the minimum duration for FC once initiated, and the subsequent transfer time limit once the air circulation is turned off to complete the transfer of the DSC into the storage module or return the DSC to the fuel handling building and refill the TC/DSC annulus with water.

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B.6 SHIELDING EVALUATION The following radiation shielding evaluation addresses the storage of a 61BTH Type 2 DSC (61BTH DSC) in a NUHOMS MATRIX (HSM-MX). It is demonstrated that the vent dose rates for storage of the 61BTH DSC are bounded by the vent dose rates for storage of either the EOS-37PTH or EOS-89BTH DSC documented in Chapter A.6. Therefore, the site dose evaluation documented in Chapter A.11 for an EOS-DSC bounds the 61BTH DSC.

It is also demonstrated that dose rates for transfer of the 61BTH DSC within the OS197 transfer cask (TC) are similar to dose rates for transfer of the EOS-89BTH DSC within the EOS-TC125 documented in Chapter 6. Therefore, the exposure estimate for transfer of the 61BTH DSC to the HSM-MX documented in Chapter B.11 is similar to the exposure estimate for transfer of the EOS-89BTH DSC to the HSM-MX documented in Chapter A.11.

The 61BTH DSC may store up to 120 irradiated stainless steel rods contained within reconstituted fuel assemblies.

The 61BTH DSC may store up to 4 failed fuel canisters (FFCs) containing failed fuel, up to 61 damaged fuel assemblies, or up to 61 intact fuel assemblies. Failed and damaged fuel shall not be present within the same DSC.

The methodology, source terms, and dose rates presented in this chapter are developed to be reasonably bounding for general licensee implementation of the EOS System.

The term reasonably bounding is quantified in Section 6.2.8. These results may be used in lieu of near-field evaluations by the general licensee, although the inputs utilized in this chapter should be evaluated for applicability by each site. Site-specific HSM-MX near-field evaluations may be performed by the general licensee to modify key input parameters.

Site dose evaluations for the HSM-MX under normal, off-normal, and accident conditions are documented in Chapter B.11. Because the arrangement and the distance to the site boundary is site-specific, compliance with 10 CFR 72.104 and 10 CFR 72.106 for the HSM-MX can only be demonstrated using a site-specific evaluation.

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B.6.2 Source Specification General BWR source term information in Section 6.2 is applicable to the OS197 and HSM-MX evaluation. Supplemental information is provided in this section.

B.6.2.1 Computer Programs Source terms are generated using the ORIGEN-ARP module of SCALE6.0 [B.6-5].

The default ge7x7-0 library is used for enrichments 1.5 wt.% U-235. Because enrichments below 1.5 wt.% U-235 are not available in the default library, library ge7x7-0-lowe is utilized for enrichments below 1.5 % U-235. This low-enrichment library is generated using the same TRITON models used by Oak Ridge National Laboratory to develop the default library, although with the enrichments reduced.

B.6.2.2 PWR and BWR Source Terms

[

]

Reasonably bounding BWR source terms are developed for the 61BTH Type 2 DSC within the OS197 TC and HSM-MX. The term reasonably bounding is quantified in Section 6.2.8. Ten HLZCs are available for the 61BTH DSC. The HLZCs are defined in the Technical Specifications, Figure 4A through Figure 4J [B.6-6].

To simplify the analysis, a hybrid HLZC is developed by selecting the hottest fuel allowed in each basket location for the ten HLZCs. The hybrid HLZC features four radial zones, as illustrated in Figure B.6-1. The total decay heat of this configuration is 42.9 kW, although the 61BTH DSC is limited to 31.2 kW. A 61BTH DSC could not be loaded in this manner because the thermal limits are exceeded, although the hybrid HLZC results in conservative source terms and dose rates. Source terms are developed for each zone of the hybrid HLZC for use in the OS197 TC and HSM-MX evaluations.

The methodology used to develop the source terms are the same as documented in Section 6.2.2. ORIGEN-ARP light element mass inputs are obtained from Table 6-6.

Burnup, enrichment, and cooling time combinations are developed to target decay heats of 0.48 kW/FA, 0.7 kW/FA, 1.2 kW/FA, and 0.54 kW/FA. These decay heats per FA correspond to the hybrid HLZC provided in Figure B.6-1. The minimum cooling time is 2 years. A constant specific power of 25 MW/MTU (4.95 MW/FA) is utilized in all source term calculations, which is a typical value for BWR fuel (see Section 3.4.6.2 of NUREG/CR-7194 [B.6-3]). The effect of specific power on source terms is discussed in Section B.6.2.9. Candidate source terms for each decay heat are summarized in Table B.6-1.

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B.6.2.4 Control Components No change to Section 6.2.4 (BWR fuel does not contain control components).

B.6.2.5 Blended Low Enriched Uranium Fuel No change to Section 6.2.5.

B.6.2.6 Reconstituted Fuel All Indicated Changes are in response to Revised RAI 6-4

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B.6.2.7 Irradiation Gases The quantity of gas generated by irradiation is 20.2 g-moles per fuel assembly, see Section T.4.6.6.4 of [B.6-4].

B.6.2.8 Justification for Reasonably Bounding Source Term Methodology No change to Section 6.2.8.

B.6.2.9 Sensitivity Study on Specific Power Due to the manner in which the design basis sources are developed, increasing the specific power has essentially no effect on OS197 TC or HSM-MX dose rates. The design basis source terms are developed to target the maximum allowed decay heat, see Table B.6-1. If the specific power is increased 20% to 30 MW/MTU, the cooling times must increase to maintain the same decay heat. The net effect is no increase in dose rate.

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A source term sensitivity study is developed for 30 MW/MTU using the same decay heat targets over the range of burnups and enrichments shown in the Technical Specification (TS) fuel qualification tables (FQTs) (TS Tables 19 and 20) [B.6-6]. The methodology described in Section B.6.2.2 is used to rank the relative strength of the 25 MW/MTU and 30 MW/MTU source terms. The results are reported in Table B.6-27 and Table B.6-28 for the OS197 TC and HSM-MX, respectively. It is observed that the dose rate perturbation is +/-1%, which is negligible.

The effect of specific power on the source term is addressed in Section 3.4.2.4 of NUREG/CR-6716 [B.6-7]. It is stated in this NUREG that specific power has little effect on neutron dose rates but may increase gamma dose rates. In the NUREG analysis, PWR source terms are developed for a burnup of 40 GWd/MTU, enrichment of 3.5%, cooling time of 5 years, and specific powers of 20 MW/MTU and 40 MW/MTU. It is stated in the NUREG that the gamma dose rate due to the 40 MW/MTU source is approximately 30% higher than the 20 MW/MTU source.

The NUREG analysis is replicated with BWR fuel in the periphery of the OS197 TC.

The conclusion is consistent with the NUREG, with an increase in gamma dose rate of 32%. This study is summarized in Table B.6-29. However, because the cooling time is treated as a fixed quantity (5 years), these two sources have different decay heats.

The decay heat of the 40 MW/MTU source is 13% higher than the 20 MW/MTU source. In the EOS methodology used to develop the design basis source terms in Section B.6.2.2, all candidate source terms in each zone have the same decay heat. If the cooling time of the 40 MW/MTU case in the NUREG study is extended to 5.75 years so that the decay heat matches the 20 MW/MTU case, the difference in dose rate between the 20 MW/MTU and 40 MW/MTU sources is within +/-1%, consistent with Table B.6-27 and Table B.6-28. This is demonstrated in the last column of Table B.6-29. Therefore, the source terms used in the UFSAR analysis have an additional decay heat constraint absent in the NUREG analysis.

Specific power would affect the dose rates only if the specific power is increased while the cooling times provided in TS Tables 19 and 20 [B.6-6] remain fixed. In this scenario, the fuel assemblies would exceed 1.2 kW/FA and 0.54 kW/FA. Dose rates would increase, but the fuel assemblies would also exceed the thermal limits, and fuel assemblies with these sources could not be stored.

The increase in FQT cooling times due to a specific power of 30 MW/MTU is small, ranging from 0 years for low burnups to approximately 0.2 years for high burnups. If the peripheral zone sources are generated for 30 MW/MTU but with FQT cooling times defined by TS Table 20 [B.6-6], the design basis OS197 TC and HSM-MX sources increase to 0.549 kW/FA and 0.566 kW/FA, respectively, see Table B.6-30.

Both decay heats exceed the peripheral zone decay heat limit of 0.54 kW/FA. Using these sources, OS197 TC dose rates increase < 2% and HSM-MX dose rates increase

< 5%. These dose rate perturbations are small and generally understood to be well-within the uncertainty of dose rate calculations and measurements.

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It is not recommended to apply dose rate scaling factors to account for a specific power of 30 MW/MTU because (1) the sensitivity analysis results in Table B.6-27 and Table B.6-28 show no effect on dose rate for higher specific power if the EOS source term development methodology is applied, and (2) if 30 MW/MTU is utilized in source term development with the existing FQT cooling times and decay heat is allowed to exceed the limits, the dose rate effect is small (< 2% for OS197 TC and < 5% for HSM-MX). In the latter scenario, fuel assemblies associated with these sources could not be stored because thermal limits are exceeded.

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B.6.4 Shielding Analysis B.6.4.1 Computer Codes MCNP5 v1.40 is used in the shielding analysis [B.6-1]. MCNP5 is a Monte Carlo transport program that allows full 3D modeling of the HSM-MX. Therefore, no geometrical approximations are necessary when developing the shielding models.

B.6.4.2 Flux-to-Dose Rate Conversion No change to Section 6.4.2.

B.6.4.3 OS197 TC Dose Rates Normal Conditions OS197 TC dose rates are computed with and without reconstituted FAs. The source terms for standard FAs are provided in Table B.6-2 through Table B.6-5. The reconstituted fuel assembly source term to be applied on the periphery for reconstituted FAs containing 5 irradiated stainless steel rods per FA (120 rods per DSC) is provided in Table B.6-21.

Dose rates are computed using mesh tallies similar to the mesh tallies utilized in the EOS-TC125 models to facilitate dose rate comparisons. To simplify the presentation, only maximum total dose rates are reported at or near the surface of the OS197 TC in Table B.6-13. This table provides dose rates for transfer and transfer peak.

Transfer dose rates correspond to the tally structure shown in Figure 6-7. Using this tally structure, the bottom and top tallies correspond to the entire bottom or top surface, and the side tallies are circumferential averages around the entire cask. The transfer peak dose rates are computed using a more refined tally structure, as indicated in Figure 6-8 through Figure 6-10. In the refined tallies, the top and bottom tallies have six annular regions, and the side tallies have 24 angular regions. While Figure 6-7 through Figure 6-10 depict the EOS-89BTH DSC within the EOS-TC125, the tally locations are similar for the OS197 TC.

Dose rates with and without reconstituted fuel assemblies are reported in Table B.6-13. Reconstituted fuel does not have a large effect on OS197 TC dose rates because the bounding source is neutron-dominated, and an increase in Co-60 activity due to irradiated stainless steel rods is largely offset by a reduction in the neutron source due to the loss of uranium-oxide rods.

The OS197 TC dose rates reported in Table B.6-13 are highly conservative due to the hybrid HLZC assumption (see Figure B.6-1). Due to the large neutron sources in each zone, approximately 40% of the dose rate at the side of the OS197 is due to fuel assemblies in the inner zones. If HLZC 1 through 10 were modeled explicitly, the dose rates would decrease compared to the hybrid HLZC.

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The maximum OS197 TC and EOS-TC125(89BTH) dose rates are compared in Table B.6-14. The dose rates computed for the OS197 TC (61BTH) are bounded by the EOS-TC125(89BTH) dose rates on the side, where the majority of operations occur. Dose rates are similar at the top of both TCs, where their magnitudes are considerably lower. Dose rates are larger for the OS197 TC on the bottom of the cask, although the cask bottom is inaccessible during decontamination and welding operations.

Occupational exposure for transfer of the EOS-89BTH DSC to the HSM-MX is provided in Table A.11-4. In the OS197 TC occupational dose assessment provided in Section B.11.2.1, EOS-TC125(89BTH) dose rates are conservatively applied for decontamination and welding operations, and OS197 TC dose rates are applied for transfer operations.

Accident Conditions The 100 m dose rate under accident conditions with only standard fuel is provided in Table B.6-15 and is 1.28 mrem/hr. When 24 reconstituted FAs are loaded into the peripheral storage locations containing 5 irradiated stainless steel rods each, the dose rate decreases slightly to 1.27 mrem/hr. These values are bounded by the maximum EOS-TC dose rate from Table 6-54.

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B.6.5.2 References B.6-1 Oak Ridge National Laboratory, MCNP/MCNPX - Monte Carlo N-Particle Transport Code System Including MCNP5 1.40 and MCNPX 2.5.0 and Data Libraries, CCC-730, RSICC Computer Code Collection, January 2006.

B.6-2 ADVANTG - An Automated Variance Reduction Parameter Generator, Oak Ridge National Laboratory, August 2015.

B.6-3 NUREG/CR-7194, Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems, US Nuclear Regulatory Commission.

B.6-4 TN Americas LLC, Updated Final Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel, Revision 18, USNRC Docket Number 72-1004, January 2019.

B.6-5 Oak Ridge National Laboratory, A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, ORNL/TM-2005/39, Version 6, SCALE, January 2009.

B.6-6 CoC 1042 Appendix A, NUHOMS EOS System Generic Technical Specifications, Amendment 2.

B.6-7 NUREG/CR-6176, Recommendations on Fuel Parameters for Standard Technical Specifications for Spent Fuel Storage Casks, US Nuclear Regulatory Commission.

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Table B.6-27 OS197 TC Side Dose Rate Sensitivity Results Side Dose Rate (mrem/hr)(1)

Zone Heat (kW/FA)

SP = 25 MW/MTU SP = 30 MW/MTU Change in Dose Rate (%)

1 0.48 27.53 27.51

-0.1%

2 0.70 104.93 104.90 0.0%

3 1.20 176.25 176.26 0.0%

4 0.54 448.24 447.82

-0.1%

Total 756.95 756.49

-0.1%

(1) Using the methodology outlined in Section B.6.2.2, each ranking dose rate represents the OS197 TC side dose rate contribution from that zone.

Table B.6-28 HSM-MX Roof Vent Dose Rate (No Vent Cover) Sensitivity Results Roof Vent Dose Rate (mrem/hr)(1)

Zone Heat (kW/FA)

SP = 25 MW/MTU SP = 30 MW/MTU Change in Dose Rate (%)

1 0.48 1.76E+04 1.76E+04 0.0%

2 0.70 2.56E+04 2.57E+04 0.1%

3 1.20 4.35E+04 4.37E+04 0.4%

4 0.54 1.98E+04 1.98E+04

-0.2%

(1) Using the methodology outlined in Section B.6.2.2, each ranking dose rate represents each fuel assembly in all 61 basket locations. In the actual configuration, the inner locations (zones 1, 2, and 3) contribute much less than the peripheral zone (zone 4) to the vent dose rates.

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Table B.6-29 BWR Sensitivity Study Similar to NUREG/CR-6716, OS197 TC SP SP = 20 MW/MTU SP = 40 MW/MTU SP = 40 MW/MTU BECT BU = 40 GWd/MTU E = 3.5%

CT =5.0 years BU = 40 GWd/MTU E = 3.5%

CT =5.0 years BU = 40 GWd/MTU E = 3.5%

CT =5.75 years Heat (kW/FA) 0.390 0.442 0.389 Emax (MeV)

Gamma Source (g/s)

Gamma Source (g/s)

Gamma Source (g/s) 5.00E-02 5.136E+14 6.046E+14 5.164E+14 1.00E-01 1.466E+14 1.756E+14 1.470E+14 2.00E-01 1.157E+14 1.430E+14 1.155E+14 3.00E-01 3.320E+13 4.055E+13 3.318E+13 4.00E-01 2.294E+13 2.868E+13 2.292E+13 6.00E-01 2.351E+14 3.061E+14 2.331E+14 8.00E-01 8.928E+14 9.809E+14 9.004E+14 1.00E+00 1.034E+14 1.299E+14 1.025E+14 1.33E+00 3.778E+13 4.385E+13 3.798E+13 1.66E+00 1.091E+13 1.360E+13 1.096E+13 2.00E+00 3.360E+11 5.218E+11 3.232E+11 2.50E+00 6.193E+11 1.114E+12 5.981E+11 3.00E+00 2.330E+10 3.700E+10 2.219E+10 4.00E+00 2.158E+09 3.415E+09 2.056E+09 5.00E+00 3.241E+06 3.344E+06 3.251E+06 6.50E+00 1.301E+06 1.342E+06 1.305E+06 8.00E+00 2.552E+05 2.633E+05 2.559E+05 1.00E+01 5.418E+04 5.590E+04 5.433E+04 Total 2.113E+15 2.469E+15 2.121E+15 Gamma Dose Rate (mrem/hr) 103.2 136.5 102.7 Change in Dose Rate (%)

32%

-0.4%

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Table B.6-30 Dose Rate Increase for 30 MW/MTU Irradiation and FQT Cooling Times System OS197 TC HSM-MX BECT BU = 62 GWd/MTU E = 3.8%

CT =7.54 years BU = 40 GWd/MTU E = 2.5%

CT = 3.86 years Heat (kW/FA) 0.549 0.566 Emax (MeV)

Gamma Source (g/s)

Gamma Source (g/s) 5.00E-02 5.761E+14 7.829E+14 1.00E-01 1.546E+14 2.331E+14 2.00E-01 1.172E+14 1.996E+14 3.00E-01 3.385E+13 5.604E+13 4.00E-01 2.201E+13 4.113E+13 6.00E-01 2.052E+14 4.656E+14 8.00E-01 1.166E+15 1.131E+15 1.00E+00 9.933E+13 1.844E+14 1.33E+00 4.496E+13 5.912E+13 1.66E+00 1.097E+13 1.986E+13 2.00E+00 1.409E+11 1.064E+12 2.50E+00 1.228E+11 2.245E+12 3.00E+00 7.622E+09 7.975E+10 4.00E+00 7.409E+08 7.376E+09 5.00E+00 1.493E+07 6.190E+06 6.50E+00 5.992E+06 2.484E+06 8.00E+00 1.176E+06 4.873E+05 1.00E+01 2.496E+05 1.035E+05 Dose Rate (mrem/hr) 456 2.07E+04 25 MW/MTU Dose Rate (mrem/hr) 448 1.98E+04 Change in Dose Rate (%)

1.7%

4.5%

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Figure B.6-2 OS197 TC MCNP Model, 61BTH DSC (x-y view)

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Figure B.6-3 OS197 TC MCNP Model, 61BTH DSC (x-z view)

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B.11.2 Occupational Dose Assessment This section provides estimates of occupational dose for the OS197 TC and ISFSI loading operations. Assumed annual occupancy times, including the anticipated maximum total hours per year for any individual, and total person-hours per year for all personnel for each radiation area during normal operation and anticipated operational occurrences, will be evaluated by the licensee in a 10 CFR 72.212 evaluation to address the site-specific ISFSI layout, inspection, and maintenance requirements. In addition, the estimated annual collective doses associated with loading operations will be addressed by the licensee in a 10 CFR 72.212 evaluation.

B.11.2.1 61BTH DSC Loading, Transfer, and Storage Operations The dose rates for the 61BTH DSC within the OS197 TC are similar to the dose rates for the EOS-89BTH DSC within the EOS-TC125 on the top and side, see the discussion in Section B.6.4.3. Therefore, the decontamination and welding dose rates for the EOS-89BTH DSC within the EOS-TC125 from Table 11-1 are conservatively applied for the OS197 TC occupational dose assessment. Transfer dose rates for the OS197 TC and HSM-MX front average dose rates are obtained from the analysis documented in Chapter B.6. Dose rates used as input for the occupational dose assessment are provided in Table B.11-1 and include reconstituted FAs containing a total of 120 irradiated stainless steel rods on the periphery. Dose rate locations around the cask are analogous to the EOS-TC125 dose rate locations illustrated in Figure 11-1.

The estimated occupational exposures to ISFSI personnel during loading, transfer, and storage operations (time and number of workers may vary depending on individual ISFSI practices) are provided in Table B.11-2. The total exposure is 2.5 person-rem.

The exposure provided is a bounding estimate. Measured exposures from typical NUHOMS System loading campaigns have been 600 mrem or lower per canister for normal operations, and exposures for the HSM-MX are expected to be similar.

Regulatory Guide 8.34 [B.11-4] is to be used to define the onsite occupational dose and monitoring requirements.

B.11.2.2 61BTH DSC Retrieval Operations Occupational exposures to ISFSI personnel during 61BTH DSC retrieval are similar to those exposures calculated for 61BTH DSC insertion. Dose rates for retrieval operations will be lower than those for insertion operations due to radioactive decay of the spent fuel inside the HSM-MX. Therefore, the dose rates for 61BTH DSC retrieval are bounded by the dose rates calculated for insertion.

B.11.2.3 Fuel Unloading Operations No change to Section 11.2.3.

All Indicated Changes are in response to Revised RAI 6-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. TBD, TBD October 2020 Revision 6 72-1042 Amendment 2 Page B.11-8 Appendix B is newly added in Amendment 2.

Table B.11-1 Occupational Dose Rates, OS197 with 61BTH DSC Dose Rate Location Averaged Segments(1)

Config.

Dose Rate (mrem/hr)

DRL1 A1-18, R11 Decon.

62 DRL2 A3-16, R10 Decon.

181 A3-16, R10 Transfer 208 DRL3 A17, R9 Decon.

98 A17, R9 Welding 113 A17, R9 Transfer DRL4 A3-11, R9 Decon.

DRL5 A1-18, R10 Transfer 164 DRL6 A17-18, R9 Transfer 14 DRL7 A17-18, R10 Transfer 22 DRL8 A2, R9 Transfer DRL9 A19, R0 Transfer 204 DRL10 A1, R10 Transfer 61 HMX-MX (HMX)

Front face surface average 48 (1) Dose rate locations analogous to Figure 11-1.

All Indicated Changes are in response to Revised RAI 6-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. TBD, TBD October 2020 Revision 6 72-1042 Amendment 2 Page B.11-9 Appendix B is newly added in Amendment 2.

Table B.11-2 Occupational Exposure, OS197 with 61BTH DSC (2 Sheets)

No.

Operation Configuration Dose Rate Location No. of People Duration (hr)

Dose Rate (mrem/hr)

Dose (person-mrem)

% of Total Dose 1

Drain neutron shield if necessary. Place an empty 61BTH DSC into an OS197 TC and prepare the OS197 TC for placement into the spent fuel pool.

N/A N/A 6

4.00 0

0 0%

2 Move the OS197 TC containing a 61BTH DSC without fuel into the spent fuel pool.

N/A N/A 6

1.50 0

0 0%

3 Remove a loaded OS197 TC from the fuel pool and place in the decontamination area.

Refill neutron shield tank if necessary.

Decon.

DRL1 2

0.25 62 31 1.3%

4 Decontaminate the OS197 TC and prepare welds.

Decon.

DRL2 2

1.75 181 634 25.8%

Decon.

DRL3 2

0.50 98 98 4.0%

5 Weld inner top cover plate.

Welding DRL3 2

0.75 113 170 6.9%

6 Vacuum dry and backfill with helium.

Welding DRL3 2

0.50 113 113 4.6%

7 Weld outer top cover plate and port covers, perform non-destructive examination.

Welding DRL3 2

0.50 113 113 4.6%

8 Drain annulus. Install OS197 TC top cover.

Ready the support skid and transfer trailer.

Transfer DRL5 1

0.50 164 82 3.3%

9 Place the OS197 TC onto the skid and trailer.

Secure the OS197 TC to the skid.

Transfer DRL2 2

0.33 208 137 5.6%

10 Install retractable roller tray (RRT).

Transfer HMX 2

2.00 48 192 7.8%

11 Transfer the OS197 TC to ISFSI.

N/A N/A 6

1.83 0

0 0%

12 Position the OS197 TC inside the loading crane (MX-LC).

Transfer HMX+DRL2 2

0.50 256 256 10.4%

13 Remove forced cooling system (if used) and install the ram cylinder assembly.

Transfer DRL9 2

0.50 204 204 8.3%

14 Remove HSM-MX door.

Transfer HMX 2

0.50 48 48 2.0%

15 Remove the OS197 TC top cover.

Transfer HMX+DRL6 2

0.67 62 83 3.4%

16 Align and dock the OS197 TC with the HSM-MX. Secure the OS197 TC to the HSM-MX.

Transfer HMX+DRL7 2

0.25 70 35 1.4%

17 Transfer the 61BTH DSC from the OS197 TC to the HSM-MX using the ram cylinder.

N/A N/A 3

0.50 0

0 0%

All Indicated Changes are in response to Revised RAI 6-4

NUHOMS EOS System Updated Final Safety Analysis Report Rev. TBD, TBD October 2020 Revision 6 72-1042 Amendment 2 Page B.11-10 Appendix B is newly added in Amendment 2.

Table B.11-2 Occupational Exposure, OS197 with 61BTH DSC (2 Sheets)

No.

Operation Configuration Dose Rate Location No. of People Duration (hr)

Dose Rate (mrem/hr)

Dose (person-mrem)

% of Total Dose 18 Disengage the ram and un-dock the OS197 TC from the HSM-MX.

Transfer HMX+DRL10 2

0.08 109 18 0.7%

19 Install HSM-MX access door. Move OS197 TC to the transfer skid for removal.

Transfer HMX 2

0.50 48 48 2.0%

20 Uninstall RRT.

Transfer HMX 2

2.00 48 192 7.8%

Total 2452 All Indicated Changes are in response to Revised RAI 6-4