Text

Enclosure 3: Responses to Request for Additional Information
	ML20190A138
Person / Time
Site:	07201042
Issue date:	06/30/2020
From:	; TN Americas LLC
To:	; Office of Nuclear Material Safety and Safeguards
Shared Package
ML20190A135	List: ML20190A136; ML20190A137; ML20190A138; ML20190A139; ML20190A140; ML20190A141; ML20190A142; ML20190A143; ML20190A144; ML20190A145; ML20198M433;
References
	E-56684
	Download: ML20190A138 (41)
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RAIs and Responses - Public Enclosure 3 to E-56684 Proprietary Information on This Page Withheld Pursuant to 10 CFR 2.390.

Page 1 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 RAI 3-2:

Justify the assumption of a temperature of 500°F used for SA-240 Type 304 stainless steel for the EOS-37PTH Dry Shielded Canister (DSC) in Appendix 3.9.1, DSC SHELL STRUCTURAL ANALYSIS.

The applicant assumed a temperature of 500°F for SA-240 Type 304 stainless steel in Appendix 3.9.1, DSC SHELL STRUCTURAL ANALYSIS (see Table 3.9.1-3), while Table 4.9.7-7 shows a maximum temperature of 717°F at the DSC shell in the heat zone loading configuration (HLZC) 10, which is the new HLZC proposed in this Amendment 2. Justify the temperature assumption in the structural analysis for the SA-240 Type 304 stainless steel. Staff is concerned that structural calculations related to pressure and thermally induced stresses may result in lower margins. Update the calculations as necessary This information is needed so that the staff may determine compliance with the regulations in 10 CFR 72.236(b).

Response to RAI 3-2:

The bounding maximum shell temperature of the EOS-37PTH DSC is 500 °F for normal and off-normal conditions. As shown in UFSAR Table 4.9.7-7, the maximum temperature of 717 °F for the DSC shell is for the bounding transfer accident condition involving accident loss of neutron shield with loss of air circulation. The evaluation of secondary stress is not required for Level D events (UFSAR Table 3-1). Thermal stress is a secondary stress and the accident condition is a Level D event.

Impact:

No change as a result of this RAI.

Page 2 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 Thermal RAIs:

RAI 4-1:

Clarify in the Safety Analysis Report (SAR) whether the Horizontal Storage Module - Matrix (HSM-MX) storage conditions and OS197 transfer cask conditions considered in Chapter B.4 reflect the different content from that found in Appendix T, Chapter T.4 of Reference [B.4-2].

Technical Specification Table 19 and Table 20 indicated that this amendment requests content in the 61BTH Type 2 DSC has the cooling period reduced to two years from the five years analyzed in reference [B.4-2], which would indicate differences in radionuclides and fission gases for a specified heat load. There was no discussion in Chapter B.4 to demonstrate that the decay heat and gases and their effect on pressure calculations (e.g., maximum internal pressure discussed in B.4.6) within the 61BTH DSC in the HSM-MX and OS197 transfer cask (and its derivatives) are as reflected in [B.4-2].

This information is needed to determine compliance with 10 CFR 72.236 (f), (l).

Response to RAI 4-1:

During the initial design review and application of the Certificate of Compliance (CoC) No. 1004 license for the 61BTH DSC, TN Americas LLC (TN) has evaluated the impact of the cooling period on the total amount of fission gases released due to irradiation. TN concluded that the 5 years cooling period results in the bounding total amount of fission gases released per fuel assembly (FA) due to irradiation for the shorter cooling periods reduced to 2 years.

The quantities of gas at reactor discharge (0-year cooling period) and 5 years cooling period are provided in Table RAI-4-1-1. As shown in Table RAI-4-1-1, the total amount of gases released per FA due to irradiation is 20.1 g-moles at reactor discharge (0-year cooling period), and increases to 20.2 g-moles after the 5 years cooling period. The longer cooling period results in a slightly higher amount of gases, primarily from the increase in helium due to alpha decay of actinides. Therefore, the amount of fission gases released per FA from the longer cooling period bounds that from the shorter cooling period. It is conservative to consider the 20.2 g-moles per FA based on 5 years cooling period as the bounding total amount of fission gases released per FA due to irradiation in the maximum pressure calculation in Section T.4.6.6.4 of the Updated Final Safety Analysis Report (UFSAR) reference [B.4-2].

As evaluated in Section B.4.4.3, the average helium temperatures calculated for the 61BTH Type 2 DSC stored in the HSM-MX are bounded by those in HSM-H for normal, off-normal, and accident conditions. Also the same OS197 transfer cask is used to transfer the 61BTH Type 2 DSC to the HSM-H or HSM-MX. Therefore, it is concluded that the maximum internal pressures listed in Tables T.4-16, T.4-20, and T.4-24 of the UFSAR reference [B.4-2] remain bounding for the 61BTH DSC loaded in the HSM-MX and OS197 transfer cask when the FA minimum cooling period is reduced to 2 years.

Page 3 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 Table RAI-4-1-1: Amount of Moles of Gases Released as Result of Irradiation for Generic BWR Fuel Assembly Reactor Discharge 5 Years Cooling (0 Year Cooling Element Period Period)

(g-moles)

H 1.232 1.230 He 0.347 0.450 N 0.976 0.976 F 0.126 0.126 Ne 0.00001 0.00001 Cl 0.034 0.034 Ar 0.000 0.000 I 0.663 0.676 Br 0.080 0.080 Kr 1.288 1.266 Xe 15.357 15.396 Total 20.1 20.2 Impact:

No change as a result of this RAI.

Page 4 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 RAI 4-2:

Clarify in the SAR the impact of increasing component temperatures (e.g., DSC) due to model uncertainties (e.g., Grid Convergence Index (GCI) corrections), for normal, off-normal, and accident conditions during transfer, especially considering that an important to safety (ITS) component is at a temperature well above from which property data are available.

It is important that corrections and uncertainties are considered to ensure temperatures are below allowable values. Although SAR Section 4.9.7.2.1.4 included the effect of the GCI correction for storage, this correction was not explicitly included in the discussions for transfer and transfer accident conditions (e.g., Table 4.9.7-4, Table 4.9.7-5, Table 4.9.7-7). This is especially relevant for temperatures of duplex stainless-steel DSCs that are above temperatures from which property data are available (see Materials RAI 8-4).

This information is needed to determine compliance with 10 CFR 72.236 (f), (l).

Response to RAI 4-2:

Based on the discussion in Updated Final Safety Analysis Report (UFSAR) Section 4.9.3.2, the GCI for the thermal model of the EOS-37PTH DSC transferred in the EOS-TC125 is 7.46 °F.

Among the various temperatures obtained, the maximum fuel cladding temperature is considered as the key simulation variable to determine the discretization error, because it has the least margin to the design limit. Therefore, the GCI of 7.46 °F is determined based on the maximum fuel cladding temperature.

The same methodology in UFSAR Section 4.9.3.2 is applied to determine the GCI based on the dry shielded canister (DSC) shell temperatures. The maximum fuel cladding temperatures are replaced by the maximum DSC shell temperatures. Table RAI 4-2-1 lists the summary of the calculation. As shown in Table RAI 4-2-1, the GCI based on DSC shell temperature is 0.61 °F, which is bounded by the GCI based on the fuel cladding temperature listed in UFSAR Section 4.9.3.2. This GCI correction is insignificant compared to the DSC shell temperature, and will not affect the performance of the DSC shell.

As shown in UFSAR Table 4.9.7-5 and Table 4.9.7-7, the EOS-37PTH DSC shell temperatures, under accident conditions, may exceed the 600 °F limit of duplex stainless steels. However, the embrittlement of the duplex stainless steel DSC shell will not occur under the accident conditions. As indicated in Section 8.2 of Certificate of Compliance (CoC) No. 1042 Safety Evaluation Report (SER) [1], the period of time the DSC shell temperatures are above the 600 °F operating limit of the material is short when compared with the time necessary for metallurgical changes to result in embrittlement of the alloy. The response to RAI 8-4 also includes the detailed evidence that embrittlement of the duplex stainless steel will not occur for the EOS-37PTH DSCs.

UFSAR Section 4.9.7.2.2.3 has been revised to include the information of the GCI for the EOS-37PTH DSC transferred in the EOS-TC125.

Page 5 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 Table RAI 4-2-1 Discretization Error of the DSC Shell Temperature of the EOS-TC125 loaded with the EOS-37PTH DSC Model Symbols Parameters Values N1 Number of elements in mesh 1 26,056,339 N2 Number of elements in mesh 2 16,666,454 N3 Number of elements in mesh 3 11,142,776 h1 (in) Representative grid size in mesh 1 0.30 h2 (in) Representative grid size in mesh 2 0.35 h3 (in) Representative grid size in mesh 3 0.40 r21 Refinement factor between mesh 2 and mesh 1 1.16 r32 Refinement factor between mesh 3 and mesh 2 1.14 1 (°F) Key simulation variable of mesh 1 495.21 2 (°F) Key simulation variable of mesh 2 495.42 3 (°F) Key simulation variable of mesh 3 495.13 avg (°F) Average key simulation variable 495.3 21=2 - 1 (°F) Difference between meshes 1 and 2 0.21 32=3 - 2 (°F) Difference between meshes 2 and 3 -0.30 p Order of accuracy 2.43 21 ext (°F) Extrapolated value 494.72 21 e a Approximate relative error 0.00043 Fs Factor of safety 1.25 21 GCI fine Non-dimensional GCI for fine grid 0.001 Unum (°F) Dimensional discretization error (GCI) 0.61

Reference:

1. Safety Evaluation Report, TN Americas LLC NUHOMS EOS Dry Spent Fuel Storage System, Docket No. 72-1042, Model No. NUHOMS EOS, Certificate of Compliance No.

1042, ML17116A278.

Impact:

UFSAR Section 4.9.7.2.2.3 has been revised as described in the response.

Page 6 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 RAI 4-3:

Provide explanation or demonstrate in the SAR that bounding thermal analyses are considered for damaged fuel and failed fuel scenarios under accident transfer conditions.

SAR Section 4.9.7.3.4 and Figure 4.9.7-7 showed that DSC and transition rail temperatures are bounding with damaged fuel (modeled as rubble) as content during accident conditions compared to the design basis condition with intact fuel at a higher decay heat. This analysis indicates that local high component temperatures are possible due to the local thermal behavior from damaged/failed fuel. However, SAR Section 4.9.7.4.4 did not perform an accident analysis with failed fuel as content but, rather, stated the accident transfer condition with intact fuel would bound the results with failed fuel because results for the intact fuel assemblies at normal/off-normal conditions bound failed fuel assembly calculations at off-normal conditions. Similar issues should be considered for the 61BTH damaged/failed fuel analyses.

This information is needed to determine compliance with 10 CFR 72.236 (f), (l).

Response to RAI 4-3:

Updated Final Safety Analysis Report (UFSAR) Section 4.9.7.3 evaluated the EOS-37PTH DSC with Heat Load Zone Configuration (HLZC) 10 loaded with 6 damaged fuel assemblies (FAs) turning into rubble during the bounding accident condition. The 6 damaged FAs, each with a heat load of 3.5 kW, are modeled as rubble with a compressed length of 44 inches at the bottom of the dry shielded canister (DSC). Due to the shorter fuel length of the damaged FAs modeled as rubble compared to the intact FAs, the heat generation rate (refer to UFSAR Section 4.4.2.3.3, Item A for definition) for the damaged FAs is higher than that of the intact FAs, which caused higher localized temperatures of the components adjacent to the damaged FAs.

Therefore, as described in UFSAR Section 4.9.7.3.4, the DSC and transition rail temperatures with the damaged FA are higher compared to the design basis results with all intact FAs during the bounding accident condition.

This accident evaluation is specifically performed for HLZC 10 with damaged FAs since the high burnup damaged FAs may undergo additional reconfiguration during a drop accident compared to their configuration during normal/off-normal conditions as indicated in UFSAR Section 4.9.6.1.5.1.

However, to bound any potential configuration for the failed FAs and their contents, they are modeled using the thermal properties of helium even for normal/off-normal conditions as noted in UFSAR Section 4.9.7.4.4. This is extremely conservative since it neglects the thermal conductivity and thermal mass of the failed FAs. In addition, similar to the accident evaluation for damaged FAs, the heat generation rates for the failed FAs are modeled over a much smaller region as noted in UFSAR Section 4.9.6.1.6.3 even for normal/off-normal conditions.

UFSAR Table 4.9.7-8 presents the results for normal/off-normal conditions based on the above assumptions and concludes that the temperatures with the intact FAs remain bounded. This behavior is primarily due to the reduction in the maximum heat load of failed FAs to 0.8 kW which is significantly lower than the 3.5 kW for intact FA at the same locations for HLZC 10.

Page 7 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 Since the failed FAs are already considered as re-configured for normal/off-normal conditions with helium thermal properties, there are no additional changes expected in their configuration during accident conditions. Because of this, a similar behavior will be observed for accident conditions as shown in UFSAR Table 4.9.7-8 wherein the maximum temperatures for intact FAs will remain bounded. Therefore, high local component temperatures observed in UFSAR Section 4.9.7.3 with damaged FAs will not be observed for the failed FAs during accident conditions.

The 61BTH DSC can accommodate up to 4 failed FAs and/or up to 12 damaged FAs with the remaining locations storing intact FAs. These damaged and failed FAs may become rubble during the worst accident conditions. To be conservative, thermal analysis was evaluated in Appendix T, Section T.4.6.9 of [2] based on the worst hypothetical condition that all damaged and failed FAs become rubble for accident conditions. Appendix T, Section T.4.6.9.4 of [2]

shows that the design basis analysis of the 61BTH Type 2 DSC with 61 intact FAs remains bounding for that with 45 intact FAs and 16 rubbles during accident conditions. As stated in UFSAR Section B.4.5.3.1, the thermal evaluation from Appendix T, Section T.4.6.9 of [2]

remains applicable for this license.

Reference

1. CoC 1042, Appendix A, NUHOMS EOS System Generic Technical Specifications, Amendment 2.

2. TN Americas LLC, Updated Final Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel, Revision 18, U.S. NRC Docket No. 72-1004.

Impact:

No change as a result of this RAI.

Page 8 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 RAI 4-4:

Clarify in the SAR how the EOS-37PTH DSC with Type 1 basket/HLZC 1 storage and transfer thermal analyses bound EOS-37PTH DSC with Type 4H basket/HLZC 1.

SAR page 4-2 and page 4-89 stated the storage and transfer thermal evaluations for EOS-37PTH Type 1 basket are applicable for the EOS-37PTH Type 4H basket. However, SAR page 4-89 stated that the time limits for transfer operations of the EOS-37PTH with Type 4H basket and HLZC 1 are reduced to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months from 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months (for the Type 1 basket/HLZC 1). It is not clear how the Type 1 basket/HLZC 1 storage and thermal analyses are bounding if a Type 4H basket/HLZC 1 requires a shorter transfer time limit.

This information is needed to determine compliance with 10 CFR 72.236 (f), (l).

Response to RAI 4-4:

As discussed in Updated Final Safety Analysis Report (UFSAR) Appendix 4.9.6.1.1, EOS-37PTH Type 4H basket has the same emissivity of steel plates and conductivity of poison material as that of Type 1 basket. In addition, the basket plates for Type 4H basket are staggered whereas it is non-staggered for Type 1 basket. As discussed in UFSAR Appendix 4.9.6.1.1, the staggering of basket plates reduces overall thermal resistance along the axial direction and thus improves heat transfer performance for the Type 4H basket compared to the Type 1 basket. Therefore, the thermal evaluations for EOS-37PTH DSC with Type 1 basket/HLZC 1 under storage and transfer conditions in Sections 4.4 and 4.5 are also applicable for EOS-37PTH DSC with Type 4H basket/HLZC 1. The time limit for transfer operations for the EOS-37PTH DSC with Type 1 basket with HLZC 1 should be applicable to Type 4H basket with HLZC 1.

However, to provide additional margin and to be consistent with other applicable HLZCs, TN Americas LLC (TN) reduces the time limit from 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for Type 4H basket with HLZC 1. Therefore, Type 4H basket with all applicable HLZCs (HLZCs 1, 4-11) have the same time limit of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months as listed in UFSAR Table 4.9.6-7 for HLZCs 4 through 6, Table 4.9.6-11 for HLZCs 7 through 9, and Table 4.9.7-6 for HLZCs 10 through 11.

Impact:

No change as a result of this RAI.

Page 9 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 RAI 4-5:

Correct the SAR and clarify the meaning and relevance of the final sentence in SAR Section 4.9.6.1.3.2 (If the maximum temperatures from the above transient analyses and the accident evaluation ).

The sentence speaks of a transient analysis even though the discussion is for a steady-state storage scenario. The sentences context, or proper location within the SAR, is uncertain and may affect proper understanding of the relevant storage/transfer system.

This information is needed to determine compliance with 10 CFR 72.236 (f), (l).

Response to RAI 4-5:

The final sentence in Updated Final Safety Analysis Report (UFSAR) Section 4.9.6.1.3.2 (If the maximum temperatures from the above transient analyses and the accident evaluation ) was incorrectly placed in UFSAR Section 4.9.6.1.3.2. The sentence has been removed from UFSAR Section 4.9.6.1.3.2 and moved to its correct location in UFSAR Section 4.9.6.1.4.1.

Impact:

UFSAR Section 4.9.6.1.3.2 and 4.9.6.1.4.1 have been revised as described in the response.

Page 10 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 RAI 4-6:

Clarify in the SAR that the internal DSC pressure within the EOS-108 transfer cask is below design basis values for off-normal and accident conditions.

SAR Section 4.9.6.3.3 C (page 4.9.6-24) notes that the DSC cavity temperature is 550 K. This is higher than the 543 K temperature for the bounding normal condition of the EOS-125 transfer cask described in SAR Section 4.9.6.1.6.4 and would appear to indicate the conditions in the EOS-108 result in higher gas cavity temperatures, and thus, higher cavity pressures, than in the EOS-125. However, no discussion was provided to demonstrate that the internal DSC pressure within the EOS-108 during off-normal and accident conditions would be below design basis values. Likewise, the effect on model uncertainties (e.g., GCI temperature corrections) should be discussed when determining internal DSC pressures.

This information is needed to determine compliance with 10 CFR 72.236 (f), (l).

Response to RAI 4-6:

UFSAR Section 4.9.6.3.3 presents the results for the thermal analysis of EOS-TC108 with EOS-37PTH DSC for bounding heat load zoning configuration (HLZC) 4 with all intact fuel assemblies (FAs) and maximum heat load of 50 kW (LC 1 from Table 4.9.6-12). A similar analysis is performed for EOS-TC125 with EOS-37PTH DSC for bounding HLZC 4 with all intact FAs in Section 4.9.6.1.4 (LC 1 from Table 4.9.6-4). Section 4.9.6.1.6 presents the thermal analysis of EOS-TC125 loaded with EOS-37PTH DSC for HLZC 6 with intact FAs along with failed fuel canisters (FFCs) and maximum heat load of 44.3 kW. Table RAI 4-6-1 lists the details for these three load cases:

Table RAI 4-6-1 Descriptions of the Bounding Normal Transfer Cases in UFSAR Section 4.9.6 Average UFSAR Load Case Description System HLZC FA type Cavity Gas Section Temp (K)

LC 1 for HLZC EOS-TC125 / 4 4 from UFSAR 4.9.6.1.4 All Intact 557 EOS-37PTH (50.0 kW)

Table 4.9.6-4 Normal, hot, LC from indoor, transient, EOS-TC125 / 6 33 Intact UFSAR Section 4.9.6.1.6 543 no air circulation, EOS-37PTH (44.3 kW) / 4 FFCs 4.9.6.1.6 120 °F ambient LC 1 from EOS-TC108 / 4 UFSAR Table 4.9.6.3.3 All Intact 550 EOS-37PTH (50.0 kW) 4.9.6-12 Page 11 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 As shown in Table RAI 4-6-1, the heat load for the load case (LC) from UFSAR Section 4.9.6.1.6 is lower when compared with the bounding LC 1 for HLZC 4 in UFSAR Section 4.9.6.3.3, and the temperature for the DSC cavity gas is also lower. The words bounding normal condition used in UFSAR Section 4.9.6.1.6.4 means that it is the bounding normal load case of the EOS-TC125 loaded with the EOS-37PTH DSC for HLZC 6. It does not mean that the LC from Section 4.9.6.1.6 with HLZC 6 is the bounding normal condition for the EOS-TC125 loaded with the EOS-37PTH DSC for all HLZCs.

To show that the EOS-TC108 temperatures are bounded by the EOS-TC125 temperatures during normal, off-normal, and accident conditions, comparisons should be made between LC 1 for HLZC 4 from UFSAR Table 4.9.6-4 and LC 1 from UFSAR Table 4.9.6-12 as these two LCs are exactly the same except for the transfer cask. As reported in UFSAR Section 4.9.6.1.4.4 B, the average helium gas temperature in the DSC cavity is 557 K for the bounding LC 1 with EOS-TC125 with HLZC 4, which is higher than 550 K reported for LC 1 with EOS-TC108 with HLZC 4 from UFSAR Section 4.9.6.3.3 C. Both of the average helium gas temperatures are lower than the design basis value of 565 K reported in Table 4-45 of UFSAR Section 4.7.

Therefore, it is concluded that the internal pressures for the EOS-37PTH DSC loaded in EOS-TC108 will remain below those for EOS-37PTH DSC loaded in EOS-TC125 during normal, off-normal and accident conditions and satisfy the design criteria limits. UFSAR Section 4.9.6.3.3 C has been updated to clarify this.

The internal pressure for the EOS-37PTH DSC includes significant conservative inputs. It is calculated based on the methodology outlined in UFSAR Section 4.7 and is applicable to all DSC systems in the UFSAR. As has been discussed above, the design basis pressures reported in UFSAR Table 4-45 for the EOS-TC125/EOS-37PTH DSC system bound those for the EOS-TC108/EOS-37PTH DSC system with sufficient margins to the design limits. Since the internal pressures remain bounded, no additional evaluations are performed in this application.

Impact:

UFSAR Section 4.9.6.3.3 C has been revised as described in the response.

Page 12 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 RAI 4-7:

Provide the time limits and the appropriateness of the time periods associated with required action A.2 and A.3 of limiting conditions for operation (LCO) 3.1.3.

a. The technical specification for LCO 3.1.3 indicated that the 61BTH Type 2 DSC time limits to ensure safe operation when completing the DSC transfer are not available. The technical specifications, which form part of the CoC approval, are to be completed prior to NRC certification to ensure there is adequate performance.

b. The appropriateness of the time limits should be discussed relative to the sensitivity of the ambient temperature during transfer, as noted in Technical Specification 5.1.2(g).

This information is needed to determine compliance with 10 CFR 72.11 and 72.236(f).

Response to RAI 4-7:

In order to provide the time limits and the appropriateness of the time periods associated with Technical Specifications (TS) required actions A.2 and A.3 of Limiting Conditions for Operation (LCO) 3.1.3, a computational fluid dynamics (CFD) evaluation of the 61BTH Type 2 Dry Shielded Canister (DSC) in the OS197 Transfer Cask (TC) during transfer operations has been added in the new UFSAR Section B.4.5.6.

UFSAR Section B.4.5.1 was updated to provide the appropriate time periods for required actions A.2 and A.3 of LCO 3.1.3. The time limits listed in Section B.4.5.1.4 and TS LCO 3.1.3 for the 61BTH Type 2 DSC with heat loads greater than 22 kW and less than or equal to 31.2 kW were reduced to ensure that sufficient time is provided to initiate the recovery actions to be consistent with the EOS-37PTH and EOS-89BTH DSCs.

A list of computer files associated with the thermal evaluations in Section B.4.5.6 is provided as , and the thermal input/output files are provided in Enclosure 10 of the submittal package.

Impact:

TS LCO 3.1.3 has been updated as described in the response.

UFSAR Section B.4.5.1 has been updated as described in the response.

UFSAR Section B.4.5.6, Tables B.4-9 through B.4-11, and Figures B.4-9 through B.4-13 have been added as described in the response.

Page 13 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 RAI 4-8:

Provide the acceptance criteria for determining an appropriate solar shield when transferring a 61BTH Type 2 DSC with ambient temperatures exceeding 100 deg F.

Technical Specification 5.1.2(g) stated that a solar shield is necessary to ensure transfer operations, but no acceptance criteria were provided to determine the geometric extent or thermal effectiveness of the solar shield.

This information is needed to determine compliance with 10 CFR 72.236 (f), (l).

Response to RAI 4-8:

The basis for using a solar shield during transfer operations when ambient temperatures exceed 100 °F (and up to the maximum off-normal ambient temperature of 117 °F) is to prevent direct insolation on the exterior surface of the transfer cask.

While a solar shield is required for transfer operations above 100 °F for the 61BTH DSC, as noted in Section 5.3.1, Item B of Technical Specification (TS) for Certificate of Compliance (CoC) No. 1004 [1], it was not required to maintain the fuel cladding temperature below the short-term limit of 752 °F as explained below. Instead, this requirement was maintained in CoC 1004 to keep consistency among operations across the various dry shielded canisters (DSCs) in that CoC.

CoC 1004 Updated Final Safety Analysis Report (UFSAR) Appendix T, Table T.4-12 of [2]

presents the maximum fuel cladding temperature in a 61BTH Type 2 DSC for horizontal transfer and storage operations at 0 °F ambient without insolation, and 100 °F ambient with insolation.

Even with such a large difference, i.e., 100 °F in ambient temperature and addition of insolation, the variations in maximum fuel cladding temperatures are 25 °F for transfer operations and 72

°F for storage operations. This shows that the fuel cladding temperature increases by 0.25 °F for the transfer operations and 0.72 °F for storage operations when the ambient temperature increases by 1 °F. Compared to the storage operations, a smaller increase in the fuel cladding temperature is observed for transfer operations due to the short term transient nature of the transfer operations.

Based on the above discussion and considering a conservative 0.72 °F increase in fuel cladding for every 1 °F increase in the ambient temperature, the maximum fuel cladding temperature during off-normal transfer operations with 117 °F ambient and insolation is 733 °F ( = 721 °F +

0.72

17 °F).

Since this temperature is less than the maximum fuel cladding temperature of 734 °F evaluated in for Appendix T, Table T.4-12 of [2] for normal operations, a solar shade is not required to maintain the fuel cladding temperature below the allowable limit. Therefore, TS 5.1.2(g) of [3] is no longer required and has been deleted.

References:

1. CoC 1004 Amendment 15, Revision 0, CoC Appendix A Technical Specifications.

Page 14 of 41

RAIs and Responses - Public Enclosure 3 to E-56684

2. TN Americas LLC, Updated Final Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel, Revision 18, U.S. NRC Docket No. 72-1004.

3. CoC 1042, Amendment 2, Appendix A, NUHOMS EOS System Generic Technical Specifications.

Impact:

TS 5.1.2(g) has been deleted as described in the response.

Page 15 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 RAI 4-9:

Demonstrate that the EOS-37PTH DSC with a heat load greater than 41.8 kW and the EOS-89BTH DSC with a heat load greater than 41.6 kW would result in component temperatures below allowable values for ambient temperatures greater than 70 deg F.

According to Technical Specification 4.5.3, an EOS-37PTH with a heat load less than 41.8 kW (or an EOS-89BTH DSC with a heat load less than 41.6 kW) is limited to a site with a maximum calculated normal average ambient temperature (corresponding to a 24-hour period) of 90 deg F. The present licensing action seeks to add EOS-37PTH and EOS-89BTH DSC with new heat load zone configurations which can have decay heats greater than 41.6 kW and 41.8 kW.

Technical Specification 4.5.3 indicates that a EOS-37PTH with a heat load greater than 41.8 kW (or an EOS-89BTH DSC with a heat load greater than 41.6 kW) can be loaded at a site with a maximum calculated average yearly temperature of 70 deg F, even if the site has a maximum calculated normal average ambient temperature corresponding to a 24-hour period of 90 deg F (or higher). The thermal analyses for heat load greater than 41.8 kW for an EOS-37PTH DSC or 41.6 kW for an EOS-89BTH DSC have not demonstrated ambient temperatures greater than 70 deg F would result in component temperatures below allowable values. It is noted that site ambient conditions for the 61BTH DSC were not included in Technical Specification 4.5.3.

This information is needed to determine compliance with 10 CFR 72.236 (f), (l).

Response to RAI 4-9:

Items 3 and 4 of Technical Specification (TS) 4.5.3 specify the normal and off-normal/accident ambient temperatures, respectively, that should be verified by a general licensee utilizing the EOS system.

Items 3.a and 3.b of TS 4.5.3 specify the normal ambient temperatures for the DSCs permitted to be stored in the EOS-HSM, whereas Item 3.c specifies the normal ambient temperature for dry shielded canisters (DSCs) permitted to be stored in the HSM-MX.

Item 4 of TS 4.5.3 specifies the off-normal/accident ambient temperature for DSCs permitted in both the EOS-HSM and the HSM-MX.

61BTH DSC 61BTH DSC is not specifically mentioned in Items 3.a or 3.b since it is not permitted to be stored in EOS-HSM. As noted in Updated Final Safety Analysis Report (UFSAR) Appendix B.1, the 61BTH DSC is only permitted to be stored in HSM-MX. Therefore, Items 3.c and 4 of TS 4.5.3 are applicable to the 61BTH DSC.

EOS-37PTH and EOS-89BTH DSCs Licensing Basis for TS 4.5.3 As indicated in Items 3.a and 3.b of TS 4.5.3, the licensing requirements for normal ambient temperature vary with the heat load only for the DSCs permitted in EOS-HSM, and need to be verified by the general licensee.

Page 16 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 The requirements laid out in this TS are based on the discussion presented in UFSAR Section 4.2 as part of Amendment 0 to CoC 1042, and have been previously reviewed and approved by the NRC for DSCs up to 50 kW.

Specifically, Section 4.3 of the Safety Evaluation Report (SER) for Amendment 0 to CoC 1042 [1] states that the ambient conditions are adequate to ensure that the fuel cladding temperatures will remain below the allowable limits as seen from the following excerpt.

TS 4.5.3, Site Specific Parameters and Analyses, requires that the system users verify the parameters and analyses mentioned in items (1)-

(4) above for applicability at their specific site.

The staff reviewed SAR Section 4.2 and confirmed that the ambient temperature ranges noted above for thermal evaluations of normal storage, off-normal storage, and off-normal transfer conditions are adequate to assure that the fuel cladding and cask component temperatures will remain below the required limits. Therefore, the staff confirmed that the system users shall verify the ambient temperature specifications in TS 4.5.3, items #3 and #4, and the related thermal analyses for applicability at their specific site. In Chapter 3 of the SER, the staff confirmed that the maximum DSC internal pressures must be less than 15, 20, and 130 psig for normal, off-normal, and accident conditions. The staffs findings here are consistent with the bases for staffs approval of CoC No. 1004 (NRC, 2014).

TN has not requested any changes to the ambient temperature requirements in TS 4.5.3 as part of application for Amendment 2 to CoC 1042.

Impact of Adding New HLZCs Application for Amendment 2 adds HLZCs 10 and 11 to the EOS-37PTH DSC when stored in the EOS-HSM. No new HLZCs are added to the EOS-89BTH DSCs in this application.

HLZCs 10 and 11 are shown in Figure 1J and Figure 1K of the Proposed TS for Amendment 2. Section 4.9.7.1 of the UFSAR presents a description of these HLZCs and concludes that HLZC 10 with a maximum heat load of 45.7 kW bounds HLZC 11 with a maximum heat load of 44.5 kW. The same licensing requirements for normal ambient temperature should apply to the EOS-37PTH DSCs with HLZCs 10 and 11. These are:

For heat loads 41.8 kW, the minimum temperature is -20 °F. The maximum calculated normal average ambient temperature corresponding to a 24-hour period is 90 °F.

For 41.8 kW < heat load 45.7 kW (HLZC 10) / 44.5 kW (HLZC 11), the minimum temperature is -20 °F. The maximum calculated average yearly temperature is 70 °F.

- In addition, for heat loads above 41.8 kW, wind deflectors shall be installed per TS 5.5.

Page 17 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 The thermal evaluation for the bounding HLZC 10 during storage in EOS-HSM is presented in Section 4.9.7.2.1 of the UFSAR, and the results are presented in Table 4.9.7-2 of the UFSAR.

As shown in UFSAR Table 4.9.7-2, the maximum temperatures are lower for HLZC 10 with the maximum heat load of 45.7 kW (Load Case 1n), compared to the design basis evaluation with HLZC 1. This is primarily due to the lower maximum heat load of 45.7 kW for HLZC 10 compared to 50 kW for HLZC 1. For HLZC 10 with the maximum heat load of 41.8 kW (Load Case 2n), the maximum temperatures are either the same or less than those of the design basis evaluation of HLZC 2. Therefore, HLZCs 10 and 11 offer greater margins to the temperature limit compared to HLZC 1 and HLZC 2.

In addition to the normal condition evaluation, an off-normal evaluation was performed for the design basis HLZC 1 with a daily average temperature of 103 °F to evaluate ambient temperatures exceeding normal conditions. The results of this evaluation are presented as Load Case # 3 in Table 4.9.5-2 of the UFSAR.

As shown in Table 4.9.5-2 of the UFSAR, the maximum fuel cladding temperature is 738 °F for this evaluation and remains below the allowable temperature limit of 1058 °F for off-normal conditions, and is also below the fuel cladding temperature limit for normal conditions of 752 °F. Since the maximum heat loads for HLZC 10 and HLZC 11 are below the 50kW used in the evaluation of HLZC 1, the maximum fuel cladding temperatures for HLZCs 10 and 11 are expected to remain below the fuel cladding limit even with temperatures exceeding normal conditions.

Reference:

1. TN Americas LLC Safety Evaluation Report, TN Americas LLC NUHOMS EOS Dry Spent Fuel Storage System, Docket No. 72-1042, Model No. NUHOMS EOS, Certificate of Compliance No. 1042, ML17116A278.

Impact:

No change as a result of this RAI.

Page 18 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 Proprietary Information on This Page Withheld Pursuant to 10 CFR 2.390.

Page 19 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 RAI 6-2:

Revise the Technical Specifications (TS) for the EOS-HSM or HSM-MX module to specify that the OS197 transfer cask (TC) will be used to transfer only the 61BTH DSC to and from the spent fuel pool to the ISFSI.

On Page 1-2 of the Updated Final Safety Analysis Report (UFSAR) revision 0 for 72-1042 Amendment 2, the applicant states: Amendment 2 of this UFSAR incorporates the 61BTH Type 2 Dry Shielded Canister (DSC) for storage in the new NUHOMS MATRIX (HSM-MX) design submitted under Amendment 1 to CoC 1042. The 61BTH Type 2 DSC is from CoC No. 1004 Amendment No. 15. The design will allow for intact, damaged, and failed fuel, the definitions of which come from CoC No. 1042 Amendment No. 1.

On the same page of the UFSAR, the applicant states: Comprehensive analyses of components to the NUH61BTH Type 2 DSC and the OS197 TC as used in the HSM-MX are provided in Appendix B [to this SAR].

All the above-quoted statements indicate that the OS197 transfer cask will be used to transfer only the 61BTH Type 2 DSC to and from the HSM-MX storage module. However, the TS for HSM-MX system does not include this requirement. The applicant needs to revise the TS for the HSM-MX storage system to explicitly state this requirement.

The staff needs this information to determine if the NUHOMS MATRIX (HSM-MX) design with the proposed amendments meet the regulatory requirements of 10 CFR 72.236(d).

Response to RAI 6-2:

Table 17 of the Technical Specifications presents the allowed system configurations for the 61BTH Type 2 DSC, which indicates that only the OS197 TC will be used to transfer the 61BTH Type 2 DSC to and from the HSM-MX storage module. Tables 15 and 16, which show the systems configurations for the EOS-37PTH DSC and EOS-89BTH DSC, respectively, also show that the OS197 is not used to transfer the EOS DSCs. Additionally, Drawing NUH-03-8001-SAR of CoC 1004 shows that the OS197 top flange opening diameter is 69.05 inches, whereas CoC 1042 Drawings EOS01-1000-SAR and EOS01-1005-SAR show that the DSC diameter is 75.5 inches. This precludes the EOS DSCs from fitting into the OS197, or any of its variations.

Impact:

No change as a result of this RAI.

Page 20 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 RAI 6-3:

Specify the other requested Boiling Water Reactor (BWR) fuels that are not listed in Table 13, BWR Fuel Assembly Design Characteristics for the 61BTH Type 2 DSC.

Section 2.3, Fuel to be stored in the 61BTH Type 2 DSC, of the proposed TS points to Table 13 of the TS for the authorized BWR fuel classes to be stored in the 61BTH DSC. However, the notes (3) to Table 13 states: Example BWR fuel designs are listed herein and are not all-inclusive. This note appears to indicate that other fuel classes beside these listed in Table 13 can be also be stored in the 61BTH DSC. In accordance to 10 CFR 72.236(a), specifications must be provided for the spent fuel to be stored in the spent fuel cask. For this reason, the application shall include specifications for all fuel assembly designs that are intended to be stored in the 61BTH Type 2 DSC.

The staff needs this information to determine if the NUHOMS MATRIX (HSM-MX) design with the proposed amendments meet the regulatory requirements of 10 CFR 72.236(a).

Response to RAI 6-3:

The BWR fuel classes that may be stored are designated in the first column of Table 13 of the Technical Specifications (TS), and Note 3 does not allow additional fuel classes not listed here.

The intention of Note 3 is to allow evaluation of fuel assemblies using the 10 CFR 72.48 process for fuel that is an allowed class per TS Table 13, column 1, but is not specifically listed in TS Table 13, column 2. All fuel to be stored would be required to meet the fuel parameters as detailed in TS Section 2.3, Fuel to be stored in the 61BTH Type 2 DSC.

Note 3 of Table 13 of the TS has been updated to provide clarity.

Impact:

TS Table 13 has been updated as described in the response.

Page 21 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 RAI 6-4:

Clarify if fuel assemblies (FAs) containing reconstitution rods are allowed to be loaded in the peripheral locations of the DSC and revise the UFSAR and/or TS to reflect these limits as necessary.

On page 1-24 of the HSM-MX UFSAR, the applicant states: Reconstituted assemblies containing up to five replacement irradiated stainless steel rods per assembly or an unlimited number of low enriched or natural uranium fuel rods or unirradiated non-fuel rods are acceptable for storage in an EOS-89BTH DSC as intact FAs.

On page 2-6 of the same UFSAR, the applicant states: Fuel assemblies are evaluated with five irradiated stainless steel rods per assembly, 40 rods per EOS-37PTH DSC, and 100 rods per EOS-89BTH DSC. The cooling time is the same as unreconstituted FAs. The reconstituted rods can be at any location in the FAs. There is no limit on the number of reconstituted FAs per DSC; the FAs containing irradiated stainless steel reconstituted rods are modeled in the inner compartments as shown in Figure 6-1 for EOS-37PTH and Figure 6-2 for EOS-89BTH of Chapter 6.

However, the staff notes that Table 1-1t of the CoC No. 1004 TS states that the maximum number of irradiated stainless steel rods in reconstituted assemblies per the 61BTH Type II DSC is 40, the maximum number of irradiated stainless steel rods per reconstituted fuel assembly is 10, and the maximum number of reconstituted assemblies per DSC with unlimited number of low enriched UO2 rods or Zr rods or Zr pellets or unirradiated stainless steel rods is 61.

The staff also notes that the TS for amendment 2 of the HSM-MX system includes: Number of RECONSTITUTED FUEL ASSEMBLIES [per the 61BTH Type II DSC [is] 61. However, information on the maximum number of irradiated stainless steel rods in reconstituted assemblies per the 61BTH Type II DSC and the maximum number of irradiated stainless steel rods per reconstituted fuel assembly is missing from both the TS and UFSAR for the HSM-MX system.

The staff needs this information to determine if the HSM-MX meet the regulatory requirements of 10 CFR 236(d).

Response to RAI 6-4:

The reconstituted fuel qualification is different between CoC 1042 (HSM-MX) and CoC 1004 (HSM-H and Standardized HSM). For consistency with the EOS-37PTH DSC and EOS-89BTH DSC, the 61BTH Type 2 DSC does not restrict the number of irradiated stainless steel rods in the CoC 1042 Technical Specifications (TS). There is no regulatory requirement for qualification to be the same for different CoCs. Where possible, CoC 1042 follows a graded approach, while CoC 1004 is an older CoC.

The 61BTH Type 2 DSC in CoC 1042 refers to the EOS-89BTH DSC for conclusions regarding reconstituted fuel. For the EOS-89BTH DSC, 5 reconstituted rods are modeled per fuel assembly (FA) and 20 inner zone reconstituted FAs are modeled (100 rods total). This configuration results in minimal change in dose rate.

Page 22 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 The treatment of reconstituted fuel in the CoC 1042 UFSAR and TS is based on follows a graded approach. Reconstituted fuel with large numbers of irradiated stainless steel rods is rare.

For the EOS-89BTH DSC, there is essentially no impact on dose rate for 5 rods per FA and 100 rods per DSC when loaded into the inner zones. The same conclusion is applicable to the 61BTH Type 2 DSC, i.e., when used in the HSM-MX, 61BTH Type 2 DSC dose rate changes are minimal for 5 rods per FA and 100 rods per DSC when loaded into the inner zones.

For the EOS-89BTH DSC, it is stated in Section 6.2.6 that The dose rate impact of allowing a larger quantity of irradiated stainless steel rods or of allowing FAs containing irradiated stainless steel rods to be placed in the peripheral zone may be addressed with a site-specific analysis.

This approach was approved in Amendment 1 for the EOS-89BTH DSC and is also applicable to the 61BTH Type 2 DSC.

To provide additional margin, UFSAR Section B.6.2.6 for the 61BTH DSC has been modified to reduce the number of irradiated stainless steel rods to 40 prior to requiring a site-specific analysis. Furthermore, the sentence in quotations above is explicitly added to UFSAR Section B.6.2.6.

Following a graded approach, limiting the number of reconstituted rods per FA or per DSC in the TS is not needed.

Impact:

UFSAR Section B.6.2.6 has been revised as described in the response.

Page 23 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 RAI 6-5:

Justify that the specific power used in the source term calculations is appropriate or revise the source terms and shielding calculations for the HSM-MX system using a more appropriate specific power.

The applicant provides the fuel depletion parameters used in generating the source terms for the BWR fuel. The staff notes that the applicant states that it used the specific power as used in NUREG/CR-7194 for BWR fuel source term calculations. However, the staff notes that NUREG/CR-7194 used three different specific powers, 25 Watts/gram (W/g), 35 W/g, and 45 W/g. It used these different specific powers for sensitivity study rather than asserting these are the actual values of the BWR core operating parameters. The staff also notes that NUREG/CR-6700 states the specific power used for BWR fuel source analyses is 35 W/g. In addition, the staff found that the Standard NUHOMS Certificate of Compliance (CoC) 1004 design used 35 W/g in its design basis source term calculations for BWR fuel. Based on the information in the cited documents, it is not clear if the specific power used by the applicant in the source term calculation is conservative. Because using a specific power lower than the actual value in the depletion calculation will give lower source terms, the staff is concerned with whether the value used by the applicant bounds all the BWR fuels to be stored in the 61BTH DSC.

The staff needs this information to determine if the NUHOMS EOS system with the requested amendments meets the regulatory requirements of 10 CFR 72.236(d).

Response to RAI 6-5:

NUREG/CR-6716 provides an assessment and recommendations on the candidate technical specification parameters for shielding analysis. The principal fuel specification parameters are defined as burnup, enrichment and cooling time.

Specific power is discussed in NUREG/CR-6716 among the secondary fuel specifications with fuel assembly type, burnable poisons, uranium mass, moderator density, and fuel cladding materials. Section 3.4.2.4 of NUREG/CR-6716 is dedicated to specific power with an assessment and a sensitivity analysis on dose rates for 20 MW/MTU, 30 MW/MTU and 40 MW/MTU, and concludes that:

The predominant effect of the specific power, near end of life, is on the short-lived fission products,

The neutron dose rate is largely unaffected by the specific power,

Gamma dose rate increases with higher specific power, however, the effect is more pronounced at short cooling times; the effect is at the level of an acceptable computer code uncertainty for greater-than-five-year cooled fuel for sources calculated with 20 MW/MTU compared to 40 MW/MTU, see Table 8 of NUREG/CR-6716.

TN has conducted a sensitivity study on specific power using actual fuel irradiation histories selected from the fuel inventories of two reactors, a 3579 megawatts thermal output and a 1912 megawatts thermal output. The average specific powers are in the range of 5 MW/FA to 5.2 MW/FA. The study encompasses cooling times as low as two years, one fuel sample actual cooling time is 1.74 years. The fuel samples are selected based on:

Page 24 of 41

RAIs and Responses - Public Enclosure 3 to E-56684

Their high burnup/enrichment (B/E) ratio, the highest B/E ratio of the samples is 14.19,

Their high burnup at end of life,

Their high specific power during the final cycle. Note that due to the fuel management scheme, a fuel assembly likely experiences higher specific power during the first two irradiation cycles when positioned in the inner region of the core and experiences lower specific power during the final cycle (s) when positioned further from the center of the core. A specific power in the range of the average specific power (defined as average power / MTU) during the final cycle would be qualified as high specific power during the final cycle.

Decay heat and radiological sources from the evaluations using the actual fuel irradiation histories are compared to those calculated with a uniform 25 MW/MTU.

Results from the TN sensitivity study are similar to the results presented in NUREG/CR-6716:

The specific power during the final irradiation cycle has a predominant effect. Fuel assemblies with specific powers in the range of 30+ MW/MTU in the first irradiation cycles and declining specific power in the last irradiation cycle have negligible decay heat and radiological source variations when compared to a uniform 25 MW/MTU.

Note that fuel assemblies with a sharper decline in specific power in the last irradiation cycle show significantly lower decay heat and gamma source when compared to a uniform 25 MW/MTU.

One fuel assembly sample with all its irradiation cycles having specific powers higher than 25 MW/MTU (in the range of 30 MW/MTU) does show higher decay heat and gamma source (about 5%) at low cooling time (two years range) when compared to a uniform 25 MW/MTU. However, there is no difference in decay heat and gamma source at 19 years cooling time, which is the actual cooling time of that fuel assembly in June 2020.

The neutron source is largely unaffected by specific power variations. Note that the neutron source for the sample with actual specific powers in all irradiation cycles higher than 25 MW/MTU is lower (4%) for all studied cooling time range (2 years to 19 years) than that computed with a uniform 25 MW/MTU.

Based on the NUREG/CR-6716 study and TN study, the specific power during the final irradiation cycle has a predominant effect; the impact on dose rates is expected to be in the range of computer code uncertainty, the neutron source is largely unaffected by the specific power variation, and the specific power during the final irradiation cycle is not expected to be in the 35 MW/MTU range due to core fuel management and fuel depletion.

As specific power is not a principal fuel specification parameter, 25 MW/MTU is an appropriate representative specific power for shielding analysis. The conservatism of a shielding analysis is primarily based on bounding sources from burn up/enrichment/cooling time combinations with B/E ratio of 16, the bounding sources are then applied to all the fuel assemblies according to the decay heat zoning requirements. Therefore, appropriate representative values can be used for the other parameters, specific power included, yet maintaining the overall conservatism of the evaluation. Note that decay heat evaluation includes appropriate uncertainties on burnup and enrichment to ensure the conservatism of the fuel selection when loading in a given heat load zone configuration.

Page 25 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 The 25 MW/MTU (25 watts/g) specific power used in the analysis is based on publicly available data in addition to NUREG/CR-7194. NUREG/CR-7194 states that a specific power of 25 MW/MTU is typical for boiling water reactor (BWR) fuel, consistent with the description in the UFSAR. Table RAI 6-5-1 provides the average life-cycle specific power for most BWR reactors currently operating (information on thermal power and number of fuel assemblies is obtained from the NRC website). All reactors operate near 25 MW/MTU per fuel assembly for an assumed fuel loading of 0.198 MTU. As described in the sensitivity study, fuel is generally burned in multiple cycles, and the power typically decreases with each cycle. The last cycle typically has the lowest power. Modeling a fixed specific power of 25 MW/MTU for all cycles is reasonable based on current domestic data for BWR reactor operation.

All fuel must meet the decay heat requirement of the basket compartment prior to loading. Fuel burned at a higher specific power would require longer cooling times to meet thermal limitations.

There are also dose rate limits on the surface of the HSM-MX defined in Section 5.1.2.c of the Technical Specifications (TS). The 25 MW/MTU specific power is a reasonable design value, and the TS limitations ensure regulatory compliance with applicable dose rate limits.

Table RAI 6-5-1 BWR Fuel Assembly Specific Power Life-Cycle Thermal Number of Specific Power Reactor Name Power (MW) Assemblies (MW/MTU)

Browns Ferry 1 3458 764 22.9 Browns Ferry 2 3458 764 22.9 Browns Ferry 3 3458 764 22.9 Brunswick 1 2923 560 26.4 Brunswick 2 2923 560 26.4 Clinton 3473 624 28.1 Columbia 3486 764 23.0 Cooper 2419 548 22.3 Dresden 2 2957 724 20.6 Dresden 3 2957 724 20.6 Duane Arnold 1912 368 26.2 Fermi 2 3486 764 23.0 FitzPatrick 2536 560 22.9 Grand Gulf 1 4408 800 27.8 Hatch 1 2804 560 25.3 Hatch 2 2804 560 25.3 Hope Creek 1 3840 764 25.4 La Salle 1 3546 764 23.4 La Salle 2 3546 764 23.4 Limerick 2 3515 764 23.2 Monticello 2004 484 20.9 Page 26 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 Life-Cycle Thermal Number of Specific Power Reactor Name Power (MW) Assemblies (MW/MTU)

Nine Mile Point 1 1850 532 17.6 Nine Mile Point 2 3988 764 26.4 Peach Bottom 2 3951 764 26.1 Peach Bottom 3 3951 764 26.1 Perry 1 3758 748 25.4 Quad Cities 1 2957 724 20.6 Quad Cities 2 2957 724 20.6 River Bend 1 3091 624 25.0 Susquehanna 1 3952 764 26.1 Susquehanna 2 3952 764 26.1 Average 24.0 Maximum 28.1 Impact:

No change as a result of this RAI.

Page 27 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 RAI 6-6:

Justify that the EOS DSC site dose analysis documented in Section A.11.3 of the UFSAR bounds the 61BTH DSC or revise the HSM-MX TS dose rate limit as necessary.

On page B.6-11 of Revision 2 of the UFSAR, the applicant states: Although a highly conservative approach is employed to compute HSM-MX dose rates using the 61BTH DSC, the dose rates are below the dose rates computed for the EOS DSCs (EOS-37PTH and EOS-89BTH), see Table B.6-16 and Table B.6-17. The TS dose rate limits in TS Section 5.1.2(c) are based upon EOS-DSC dose rates. On page B.11-5, the applicant states: The vent dose rates for an HSM-MX containing an EOS-DSC bound the vent dose rates for the 61BTH DSC, see the discussion in Section B.6.4.4. Therefore, the EOS DSC site dose analysis documented in Section A.11.3 may be used to bound the 61BTH DSC. In addition, the applicant states on page B.11-4 of the USFAR: No change to Section A.11.2.5, as the 61BTH DSC is bounded by the EOS-DSC within the HSM-MX. However, the comparison shown in Table B.6-17 of the UFSAR shows that the dose rate of 61BTH is 30% more than the dose rate of EOS-DSC at the door centerline line of the HSM-MX. As such, it is not clear to the staff if the dose rate for the HSM-MX loaded with the 61BTH DSCs is still bounded by the dose rate of the HSM-MX loaded with EOS DSC.

The staff needs this information to determine if the HSM-MX system meets the regulatory requirements of 10 CFR 72.236(d).

Response to RAI 6-6:

The door centerline dose rate is not an input to the site dose analysis. The site dose input is the average dose rate (or flux) on the front and roof of the HSM-MX. From UFSAR Table B.6-17, the average front dose rate is 40.7 mrem/hr for the 61BTH DSC versus 51.1 mrem/hr for the EOS-DSC. Furthermore, the average roof dose rate is 104 mrem/hr for the 61BTH DSC vs. 206 mrem/hr for the EOS-DSC. Skyshine from the roof dominates the site dose rate at long distances, and the roof dose rate for the 61BTH DSC is approximately half the EOS-DSC roof dose rate. Therefore, site dose rates for the 61BTH DSC would be significantly less than the EOS-DSC.

Furthermore, it is stated in Section B.6.4.4 that the door centerline dose rate is approximately 30% larger for the 61BTH DSC compared to the EOS-DSC, although both dose rates are negligibly small (2.57 mrem/hr for the 61BTH DSC versus 1.97 mrem/hr for the EOS-DSC).

The dose rate difference is 0.6 mrem/hr, which is a local effect at the door centerline with no consequence for site dose.

Impact:

No change as a result of this RAI.

Page 28 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 Editorial SH-RAI-1:

Revise the UFSAR to clearly state that damaged or failed BWR fuel is a new content in amendment 2 to the EOS HSM-MX system design and can be stored only in the 61BTH dry shielded canister (DSC).

On page 1-2 of the UFSAR, the applicant states: Amendment 2 of this UFSAR incorporates the 61BTH Type 2 dry storage canister (DSC) for storage in the new NUHOMS MATRIX (HSM-MX) design submitted under Amendment 1 to CoC 1042. The 61BTH Type 2 DSC is from CoC No. 1004 Amendment No. 15. The design will allow for intact, damaged, and failed fuel, the definitions of which come from CoC No. 1042 Amendment No. 1. However, the staff notes that damaged and failed BWR fuels are not part of the authorized contents of CoC No. 1042 Amendments No. 0 and 1. EOS-89BTH is allowed to store only intact BWR fuel. The intact, damaged, and failed BWR fuel is to be stored only in the 61BTH Type 2 DSC, not to be stored in the EOS-89BTH DSC with this amendment.

Response to Editorial SH-RAI-1:

Chapter 1 Section 1 of the Updated Final Safety Analysis Report (UFSAR) has been revised to indicate that the EOS-89BTH DSC is allowed to store only intact boiling water reactor (BWR) fuel.

Impact:

UFSAR Chapter 1 Section 1 has been revised as described in the response.

Page 29 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 Materials RAIs:

RAI 8-1:

Clarify the design criteria and American Society of Mechanical Engineers (ASME) code alternative information incorporated by reference from the 1004 FSAR Revision 18.

The EOS UFSAR section B.8.2.1 references Table R.3.1-2. No such table is included in 72-1004 UFSAR Appendix R. The staff notes that Appendix R is for the NUHOMS Horizontal Storage Module (HSM) Model 152. It appears that the callout should be to Table T.3.1-2 which has alternatives to the ASME code for the 61BTH DSC. This would be consistent with the EOS UFSAR section B.8.2.1 call out to Table T.3.1-3 that has the code alternatives to the DSC basket. Note that the Alternatives to the ASME Code are now in Section 4.2.4 of the Technical Specifications.

None of the tables referenced in B.8.2.1 apply to the OS197. For the OS197, the EOS UFSAR should include Chapter 3 Section 3.2.5.3 of the 72-1004 UFSAR and Chapter 4 Section 4.9 and Table 4.9-1 which documents the ASME Code Exceptions List for the Transfer Cask.

This information is needed to determine compliance with 10 CFR 72.236(b).

Response to RAI 8-1:

The typographical error in the Updated Safety Analysis Report (UFSAR) Section B.8.2.1 reference to Appendix T was corrected to reference the proper sections of applicable codes and standards. ASME code alternatives for the 61BTH Type 2 dry shielded canister (DSC) were also added as a reference.

References to UFSAR Sections 3.2.5.3 and 4.9, and Table 4.9-1 of Certificate of Compliance No. 1004 UFSAR Revision 18 for the OS197 were added to Section B.8.2.1.

Impact:

UFSAR Section B.8.2.1 has been revised as described in the response.

Page 30 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 RAI 8-2:

Revise UFSAR Section B.8.2.2 to be consistent with UFSAR Section B.3.3.2 (as revised in response to Materials RSI-2) to provide correct references to the OS197 TC material specifications and properties.

Appendix T Section T.3.3 of the 72-1004 UFSAR referenced in EOS UFSAR Section B.8.2.2 addresses the NUHOMS 61BTH DSC, Zircaloy and, to a limited extent, the concrete of the HSM/HSM-H. Appendix T Section T.3.3 in Reference B.8-2 does not address the OS197 Transfer Cask (TC). The 72-1004 UFSAR (reference B.8-2) Appendix T Section T.3.3 Mechanical Properties of Materials calls out Table T.3.6-3 Mechanical Properties of Materials which has entries for Stainless Steel ASME SA-240 Type 304 and Carbon Steel ASME SA-36.

The OS197 TC drawings are in Appendix E.3 of the 72-1004 UFSAR. There are many materials in the ITS structures, systems, and components (SSCs) for the OS197 TC that are not common with the 61BTH that are not included in Table T.3.6-3 in the CoC 72-1004 UFSAR.

This information is needed to determine compliance with 10 CFR 72.236(b).

Response to RAI 8-2:

The structural materials and respective material properties for the OS197 from UFSAR Section B.3.3.2 have been moved to Section B.8.2.2, and additional important-to-safety (ITS) materials have been added based on the drawings in Appendix E.3. Additionally, Section B.8.2.2 has been updated to remove the reference to the OS197 TC as being discussed in Certificate of Compliance (CoC) No. 1004 Updated Final Safety Analysis Report (UFSAR) Section T.3.3.

The upper trunnion is fabricated from American Society of Mechanical Engineers (ASME) SA-533 Gr. B, Cl. 2 or SA-508, Cl. 3A (row 8). ASME SA-533 Gr. B is not listed in Table 8.1.3; however, all properties are completely bounded by ASME SA-508, Cl. 3A, which is listed in UFSAR Table 8.1.3, row 8. The lower trunnion is constructed of SA-479 Type 304 bar stock, whose properties are essentially the same as the SA-240 Type 304 plate as provided in Table 8.1.3, row 1. The lower trunnion sleeves may be fabricated from ASME SA-516 Gr. 70 (row 4) or SA-508 Cl. 1A, the stress intensity and yield strength are not bounded by the ASME SA-516 Gr. 70; however, the OS197 lower trunnion analysis assumes that the lower trunnion sleeve is SA-182 F304, whose properties are also essentially identical to SA-240 Type 304, as provided in Table 8.1.3, row 1. The properties for the SA-240 Type 304 are conservative for both the SA-516 Gr. 70 and the SA-508 Cl. 1A. Additionally, the model uses the modulus of elasticity for 304 stainless steel (SS) for the finite element analysis, which is lower than the modulus of elasticity of carbon steel.

Impact:

UFSAR Sections B.3.3.2 and B.8.2.2 have been revised as described in the response.

Page 31 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 RAI 8-3:

Provide the range of operating temperatures for the HSM-MX components included in Drawing MX-5001-SAR and verify that these materials are suitable for the expected service temperatures under normal, off-normal and accident conditions.

These components include carbon steels, stainless steels and high strength alloy steel fasteners with the American Society of Testing and Materials (ASTM) material specifications.

The staff note that ASTM specifications do not include allowable temperature ranges for use nor do they provide information on mechanical properties as a function of temperature. The expected performance of the carbon and stainless steel materials may be estimated from information available from ASME B&PV code Section II Part D when ASME B&PV code Section II has incorporated ASTM Standards and specified material properties as a function of temperature. Relevant information on high strength steel bolting identified Drawing MX-5001-SAR is not available in the ASME B&PV code Section II Part D. Limited studies published by the National Institute of Standards and Technology (NIST) show that these bolting materials have reduced strength at elevated temperatures (Weigand el al., 2018).

This information is needed to determine compliance with 10 CFR 72.236(b).

Reference Weigand, J.M., R. Peixoto, L.C.M. Vieira Jr., J.A. Main, and M. Seif, An Empirical Component-Based Model for High-Strength Bolts at Elevated Temperatures, Journal of Construction Steel Research, Vol. 147, pp.87-102, 2018 Response to RAI 8-3:

The maximum temperatures of fuel cladding of the HSM-MX loaded with the EOS-37PTH DSC (the bounding condition, as it contains the highest heat load) for all the load cases are described in Table A.4-16 of the UFSAR.

For normal hot storage condition, the maximum fuel cladding temperature is 704 °F, which is within the temperature limit of 752 °F.

For off-normal hot storage condition, the maximum fuel cladding temperature is 685 °F, which is within the temperature limit of 1058 °F.

For accident blocked inlet vents condition, the maximum fuel cladding temperature after 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months of blocked vent accident is 770 °F, which is within the temperature limit of 1058 °F.

As shown above, the maximum fuel cladding and concrete temperatures are compared with their respective temperature limits from NUREG-1536 [1], and all fall within the acceptable range.

Page 32 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 Of all the important-to-safety (ITS) components listed in Drawing MX01-5000-SAR, Revision 0C Bill of Materials (BOM), the applicability and maximum temperature limits for American Society of Mechanical Engineers (ASME) SA-193 Grade B7, ASME SA-240 Type 304 and 316, and ASME SA-36 provided in ASME Section II Part D, Table 3 for Section III materials is used to estimate the expected performance of the similar ASTM specifications for ASTM A193 Grade B7, ASTM A240 Type 304 or 316 Stainless Steels, and ASTM A36 Carbon steel. Note that Revision 0C to this drawing was submitted as part of the RAI response associated with Amendment 1 (Accession Number ML19176A315), which occurred after Revision 0 of Amendment 2 was submitted. Although it has no effect on the contents of this response, Revision 0 to Drawing MX01-5000-SAR is incorporated into Revision 3 of the UFSAR, submitted to the NRC on June 17, 2020 and is available by way of Accession Number ML20169A600.

ASTM A193 Grade B7 is specified as the bolting material for the horizontal storage module (HSM) door embedments, axial retainer cover, DSC front support, retractable roller tray (RRT) cover, inspection port cover, cask restraint embedment, shield wall, and heat shield fasteners.

Materials specified as ASTM A240 Type 304 or 316 include the DSC support plates, heat shields (roof, floor, and walls), dose reduction hardware mounting screws, and bird screens. All materials specified as ASTM A193 Grade B7, ASTM A240 Type 304 or 316 were determined from ASME Section II, Part D Code, Table 3 for Section III materials [2] to have an applicability and maximum temperature limit of 800 °F. The maximum temperature limit of 800 °F for these components will not be exceeded because the maximum fuel cladding temperature is limited to 770 °F even in an accident condition as described above.

Another frequently used material from MX-5000-SAR is ASTM A36 Carbon Steel. ASTM A36 Carbon Steel items from the BOM include the door hatch and plate embedments, DSC support spacers, inspection port cover plate, RRT cover plate, and various dose reduction hardware.

Using ASME Section II, Part D Code, Table 3 for Section III materials [2], it was found that the maximum temperature limit for ASME SA-36 Carbon Steel is 650 °F for bar form, and 700 °F for plate form. In this case, the maximum temperature limit is not bounded by the maximum fuel cladding temperature, so expected component temperatures were compared individually against their respective maximum temperature limits. Figures A.4-21 through A.4-23 of the UFSAR illustrate the temperature profiles for the HSM-MX loaded with EOS-37PTH DSC at Bounding Normal Hot Storage Condition, Bounding Off-Normal Hot Storage Condition, and Accident Blocked Inlet Vents Storage Condition for 32 Hours for HLZC 7, which is rated for the highest maximum heat load of 50.00 kW. Based on Figure A.4-23(i) for Accident Blocked Inlet Vent Storage Condition for 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months , it is conservatively estimated that the high temperature for an ASTM A36 component, regardless of bar or plate form, is located at the door plate embedment, which may reach a maximum temperature of 291 °F - a value that is significantly lower than the applicability and maximum temperature limit of 700 °F for ASME SA-36 steel plate as prescribed by the ASME code. Based on these temperature profiles, the maximum accident condition temperatures for each carbon steel component was found to be well below the maximum temperature limit of 650 °F for bar, and 700 °F for plate, therefore demonstrating that the ASTM A36 carbon steel components will be operating within their established temperature limits.

Page 33 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 Structural steel components are fabricated from ASTM A588 for the Axial Retainer bar, and ASTM A572 Grade 50 and ASTM A992 Grade 50 for the Rear DSC Supports (note: the Front DSC Supports are fabricated from ASTM A240 type 304 or 316 and bolted with ASTM A193 Grade B7 materials, which are discussed earlier in this response). The design of these components is based on American Institute of Steel Construction, Manual of Steel Construction (AISC) According to Table A-4.2.1 of [4], which provides ratios for the mechanical properties at elevated temperatures, structural steel is expected to maintain its yield and ultimate strength up to 750 °F. As part of a response to Amendment 1 RAI 8-6 [3], it was demonstrated that the rate of reduction applied to the mechanical properties for these materials presented in UFSAR Tables A.8-2, A.8-3 and A.8-4 are conservative when compared with the AISC ratios. Based on Figure A.4-23(i) of the UFSAR, axial retainer bar is expected to see higher temperatures than the Rear DSC Support. It is not explicitly modeled; however, the axial retainer is assembled through the HSM-MX door to be within 0.1875 inch of the DSC bottom face as described in Section A.3.4.4.2.8. The temperature of the bottom face of the DSC during an accident condition is estimated to be 425 °F, so this temperature is conservatively applied to the axial retainer bar. Since the maximum temperature for these steels is expected to be far below the limit of 750 °F, both ASTM A588, ASTM A572 Grade 50 and ASTM A992 Grade 50 are all expected to be suitable for use in the HSM-MX.

The DSC Rear Support Fastener is fabricated from ASTM F3125 Grade A490 Steel. This RAI text also suggests that Limited studies published by the National Institute of Standards and Technology (NIST) show that these bolting materials have reduced strength at elevated temperatures (Weigand et al., 2018). However, Weigand et al. also states the following: For both the A325 and A490 bolts, the reference load was only slightly degraded for temperatures up to 200 °C (392 °F), with more significant degradation at 400 °C (752 °F) and above. These trends are consistent with observations by Yu (2006), who also found that high-strength bolts did not experience significant degradation in strength at temperatures less than 300 °C. [5] The temperatures for the top portion of the DSC contacting the rear supports from the DSC temperature profile in Figure A.4-23(g) were conservatively applied to the rear DSC support fasteners, therefore a temperature of 335 °F was assumed for these components. This temperature is below the temperature where degradation is expected to occur, and the accident condition is only a short term condition. Therefore, it is expected that the mechanical properties for the ASTM F3125 Grade A490 steel will be maintained in the HSM-MX.

The Axial Retainer Cover is fabricated from ASTM A514 steel and the Cask Restraint Embedment is fabricated from ASTM A193 Grade B8 CL.2, ASTM A193 Grade B8M CL.2, or ASTM A193 Grade B7. Both of these components are located outside of the door and will therefore not experience high temperatures because they are exposed to atmosphere.

References:

1. NUREG-1536, Revision 1, Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility.

2. American Society of Mechanical Engineers, Boiler & Pressure Vessel Code,Section II, Part D, 2010 Edition with 2011a Addenda.

3. E-54363, Application for Amendment 1 to NUHOMS EOS Certificate of Compliance No.

1042, Revision 5 - Revised Responses to Request for Additional Information Responses, dated June 19, 2020 (NRC ADAMS Accession number ML19176A315).

Page 34 of 41

RAIs and Responses - Public Enclosure 3 to E-56684

4. ANSI/AISC 360-10 Specification for Structural Steel Buildings, June 22, 20010.

5. Weigand, Jonathan M. Weigand; Rafaela Peixoto, Luiz Carlos Marcos Vieira Junio, Joseph A. Main; Mina Seifa; An Empirical Component-Based Model for High-Strength Bolts at Elevated Temperatures, National Institute of Standards and Technology, March 1, 2018.

Impact:

No change as a result of this RAI.

Page 35 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 RAI 8-4:

Provide the following information with respect to the analysis shown in Tables 4-29, 4.9.6-6, 4.9.7-5 and 4.9.7-7 where the maximum DSC shell temperature is greater than 600°F.

1. Provide mechanical properties for the SA-240/SA-479 Type 2205 / SA-182 Gr F60 (UFSAR Table 8-7) and SA-240 UNS S31803 / SA-182 Gr F51 (UFSAR Table 8-8) at temperatures that encompass the use of these materials under the conditions identified in UFSAR Table 4.9.7-7.

2. Explain the recovery actions for the analysis shown in Table 4.9.7-7 where the maximum DSC shell temp could exceed 600°F. The maximum DSC shell temperature of 717°F based on a steady state analysis for the accident condition with the EOS-37PTH DSC and HLZC 10 with 6 damaged fuel assemblies using shown in Table 4.9.7-7 is 117°F greater than the American Society of Mechanical Engineers (ASME) code allowable for a Unified Numbering System (UNS) S31803 duplex stainless steel using ASME Code Case N-635-1. Previous analysis show that significant alteration of the microstructure and mechanical properties can occur when duplex stainless steel are exposed to temperatures above 600°F (Weng et al. 2003; Taveres et al., 2005; Della Rovere et al.,

2013).

This information is needed to determine compliance with 10 CFR 72.236(b).

References Della-Rovere, C.A., F.S. Santos, R. Silva, C.A.C. Souza, and S.E. Kuri, Influence of Long-Term Low-Temperature Aging on the Microhardness and Corrosion Properties of Duplex Stainless Steel, Corrosion Science, Vol. 68, pp. 84-90, 2013.

Taveres, S.S.M., V.F. Terra, P. De Lima Neto, and D.E Matos, Corrosion Resistance Evaluation of the UNS S31803 Duplex Stainless Steels Aged at Low Temperatures (350 to 550 °C) using DLEPR Tests, Journal of Materials Science, Vol. 40, pp. 4023-4028, 2005.

Weng, L.W., T.H. Chen, and J.R. Yong, The High-Temperature and Low-Temperature Aging Embrittlement in a 2205 Duplex Stainless Steel, Bulletin of College Engineering, No. 89, pp. 45-61, 2003.

Response to RAI 8-4:

Response to Question 1 UFSAR Tables 4-29, 4.9.6-6, 4.9.7-5 and 4.9.7-7 provide results for Load Case 5 (LC5), which evaluates the thermal performance of the EOS-37PTH DSC when subjected to a maximum heat load of 45.7 kW per HLZC 10 under bounding accident conditions. The bounding thermal accident condition combines both the loss of water in the neutron shield and the loss of the air circulation system. Loss of the neutron shield may occur due to a transfer cask drop or a missile impact. In either accident, the air circulation system could become disengaged from the transfer cask. Loss of air circulation would result in dry shielded canister (DSC) temperatures presented in the referenced tables. At the time a loss of neutron shield accident occurs, either as a result of the drop accident or tornado missile accident, the DSC is at normal operating temperatures.

Therefore, the structural analysis evaluates the material properties for the DSC at 500 °F. Per Page 36 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 structural evaluations discussed in Section 3.5 of NUREG-1536: Structures important to safety are not required to survive accidents to the extent that they remain suited for use for the life of the cask system without inspection, repair, or replacement. If the service life of structures important to safety may be degraded by accident-level conditions, there must be SAR commitments and procedures for determining and correcting the degradation and performing other acceptable remedial action.

Therefore, demonstration of structural adequacy of the DSC after the accident conditions is not required. After such accident conditions, the DSC is only subject to dead weight and internal pressure. Furthermore, as addressed in the response to the second question, the DSC is not likely to be subject to handling loads until the temperature is reduced to within the normal condition limits. Therefore, no stress analysis is performed for the DSC at the temperatures presented in these load cases, and the material properties at temperatures reported in Table 4.9.7-7 are not required.

Response to Question 2 TN Americas LLC (TN) addressed the concern about high temperature embrittlement in the Technical Specification code alternative to ASME NB-2121. The embrittlement is a function of time and temperature. This concern was addressed in the SER to CoC 1042 Amendment 0, Section 8.2 as quoted below:

As discussed in detail below, duplex stainless steels can be susceptible to embrittlement under certain thermal exposures. The applicant showed in SAR Figure 4-16 that, under accident conditions, the shell temperatures for the EOS-37PTH DSC may exceed the 600 °F limit. However, TS Section 4.4.4 states that accident conditions would only exceed the limit for durations too short to cause embrittlement.

The staff reviewed the thermal stability of duplex stainless steels at temperatures consistent with the reported hypothetical accidents involving blocked vents to ensure that the duplex stainless steels would not be susceptible to embrittlement. Results of testing reported by Chen et al. (2002) and time-temperature-transformation information published by Outokumpu (2014) indicate that the formation of the embrittling sigma phase in duplex stainless steels occurs rapidly at temperatures of 1,652 °F. Weng et al. (2003) showed that exposure temperatures of 752 °F for 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months resulted in decreases in the measured Charpy V-notch impact energy of Type 2205 duplex stainless steels. Alteration of mechanical properties, including increased hardness and decreased impact strength, as a result of spinodal decomposition of the ferrite phase (often termed 885 °F embrittlement) can occur at lower temperatures. However, the time required for meaningful changes to material toughness or corrosion resistance at a temperature of 662

°F is more than 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months (Taveres et al., 2005; Della Rovere et al., 2013). The staff noted that TS Section 5.1.3 requires either daily visual inspections of HSM inlets and outlets or daily HSM temperature measurements to ensure that such prolonged periods of elevated temperatures do not occur. Therefore, the staff concludes that embrittlement of the S32203 or Type 2205 duplex stainless steel DSC shell will not occur under the hypothesized accident conditions because the period of time the DSC shell temperatures are above the materials operating limit of 600 °F is short when compared with the time necessary for metallurgical changes to result in embrittlement of the alloy.

Page 37 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 For DSCs with heat load zone configurations (HLZCs) with a total heat load greater than 36.35kW, UFSAR Table 4.9.7-7 reports the maximum accident temperature 717 °F of EOS-TC125 with EOS-37PTH DSC at 50kW. This is significantly below the 885 °F, which is the peak of the lower temperature range where duplex stainless steels can become brittle rapidly.

The maximum accident temperature of 717 °F is also below the 752 °F where steady state exposure for 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months would adversely affect the ductility, and this is only the maximum of a transient, not a steady state condition.

In the transient condition, after the initial 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , the heat up rate for the shell is 10 °F over 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Once the DSC shell reaches the 662 °F threshold, it would reach its maximum DSC shell temperature of 717 °F in an additional 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months . This same rate of temperature increase can be applied for the temperature decrease, for a total of 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months spent above 662 °F. This is significantly below the 500 hour0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months threshold for significant changes to material toughness at 662 °F. As such, embrittlement is not a concern for the DSC shell based on the limited exposure resulting from the accident condition.

In the case of a loss of neutron shield and air circulation, the corrective actions provided in UFSAR Section 12.3.1 for the accident drop case, and Section 12.3.4 for the tornado missile accident would be engaged. The DSC inspection for damage would include an evaluation of the accident specifics such as the decay heat load, the ambient temperature, the time from loss of cooling or neutron shield water to the time the cooling or water are re-introduced, etc. The transfer cask and DSC would be moved only after these evaluations determined it was safe to do so. Any further DSC operations would take place in the plant fuel building decontamination area and spent fuel pool after recovery of the EOS-TC.

The specific detailed actions to be taken are unique to the accident circumstances, and will be driven by the corrective action and engineering programs of the licensee and the certificate holder. The direction provided in UFSAR Sections 12.3.1 and 12.3.4 is revised to add further discussion on the DSC inspection after an accident.

Impact:

UFSAR Sections 12.3.1 and 12.3.4 have been revised as described in the response.

Page 38 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 Operating Procedures RAI:

RAI 9-1:

Revise the Operating Procedures to include requirements for measuring the dose rates around the transfer cask.

The applicant developed operating procedures for loading the 61BTH DSC into the HSM-MX module. However, the dose rate measurement action, as listed in step 19 of the Operating Procedures of the Standard NUHOMS, CoC 72-1004, is not included in the Operating Procedures for the HSM-MX for loading the 61BTH DSC. It is imperative to include the dose rate measure requirement in the Operating Procedures to ensure grossly misloaded DSCs are not loaded into the ISFSI to avoid violation of the 10 CFR 72.104 and 72.106 regulations.

The staff needs this information to determine that the HSM-MX system meets the regulatory requirements of 72.234(f) and 72.236(d).

Response to RAI 9-1:

The operating procedures in Updated Final Safety Analysis Report (UFSAR) Chapter B.9 for loading the 61BTH dry shielded canister (DSC) are developed to be consistent with the previously-approved EOS methodology. UFSAR Chapters 9 and A.9 do not require dose rate measurements for the EOS-TC when loaded with the EOS-DSC, although dose rate measurements are constantly taken by health physics and field service operators to ensure dose rates are ALARA. Since the 61BTH DSC and the EOS-89BTH DSC may be loaded in the same HSM-MX array, the loading operations between the two DSCs are meant to be maintained consistent to the extent possible.

Use of the dose rate measurement during welding/decontamination performed is a redundant function to the radiation protection monitoring performed at the independent spent fuel storage installation (ISFSI). Section 5.1.2c of the Technical Specifications (TS) provides dose rate limits for the HSM-MX at the front face, door centerline, and exterior side wall of the HSM-MX as follows:

i. 55 mrem/hr average over the front face, ii. 10 mrem/hr at the door centerline, and iii. 5 mrem/hr average at the exterior side wall of the HSM-MX monolith If the dose rates at any of these three locations are found to exceed the dose rate limits, as part of the corrective actions, the user must administratively verify that the correct fuel was loaded.

An analysis would also be performed to determine that the placement of the as-loaded DSC at the ISFSI will not cause the ISFSI to exceed the radiation exposure limits of 10 CFR Part 20 and 10 CFR Part 72 and/or if additional shielding is required to ensure exposure limits are not exceeded. Note that the average front face dose rate (TS 5.1.2c,i) was revised to bound the average dose rate reported in Table A.6-2a.

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RAIs and Responses - Public Enclosure 3 to E-56684 Additionally, dose rates are monitored at the site boundary to ensure that the 10 CFR 72.104 and 10 CFR 72.106 annual dose limits to an individual are not exceeded. In both cases, as discussed in UFSAR Sections B.11.3.1 and B.11.3.2, the maximum normal and accident exposures are not explicitly analyzed for the 61BTH DSC, but instead, are considered to be bounded by the analysis presented for the EOS-DSC in the HSM-MX. The analyses and results in the UFSAR are intended to provide high estimates of dose rates for generic ISFSI layouts.

The site-specific site dose evaluations for the actual ISFSI must consider the type and number of storage units, layout, characteristics of the irradiated fuel to be stored, site characteristics (e.g., berms, distance to the controlled area boundary), and reactor operations at the site in order to demonstrate compliance with 10 CFR 72.104 and 10 CFR 72.106. For a 2x11 array, at a site boundary of 370 m from the ISFSI, the normal annual dose is expected to be below the 10 CFR 72.104 limit of 25 mrem. At a distance of 200 m from the ISFSI, the accident dose is significantly less than the 10 CFR 72.106 limit of 5 rem. Actual site boundary dose rates are expected to be significantly less than the UFSAR dose rates to further ensure that 10 CFR 72.104 and 10 CFR 72.106 are met. This radiological monitoring program ensures that any misloading would be detected at the ISFSI, and that corrective actions are in place in the event that a misloading were to occur.

Impact:

Technical Specifications 5.1.2 has been revised as described in the response.

Page 40 of 41

RAIs and Responses - Public Enclosure 3 to E-56684 Radiation Protection RAI:

RAI 11-1:

Provide operating procedures with estimated time for each step for using the OS197 TC to transport the 61BTH to the HSM-MX storage module and corresponding dose rate estimates and radiation protection plan.

In Section B.1.2.3.1 of Revision 0 of the UFSAR for CoC 72-1042 Amendment 2 application, the applicant states: The primary operations for loading fuels into the NUHOMS 61BTH Type 2 DSC, moving the loaded OS197 TC to ISFSI, and transferring the NUHOMS 61BTH Type 2 DSC to the HSM-MX is same as described in Section A.1.2.3.1. On page B.6-1 of the same UFSAR, the applicant states: 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 EOS-89BTH DSC to the HSM-MX documented in Chapter A.11 may be applied to the 61BTH DSC. However, the staff notes from Table B.6-14 of 72-1042 Amendment 2 UFSAR that the dose rates at the bottom of the OS197 TC containing the 61BTH DSC is twice as much as the EOS-TC125 containing EOS-89BTH DSC. Based on these data the estimated exposure for transfer of the EOS-89BTH DSC to the HSM-MX documented in Chapter A.11 may not be applied to the 61BTH DSC.

The staff needs this information to proceed with its review to determine that the NUHOMS MATRIX (HSM-MX) loaded with the 61BTH canister meets the regulatory requirements of 10 CFR 72.236(d).

Response to RAI 11-1:

The bottom dose rate contributes little to operational dose, which is why a more explicit operational exposure estimate was not originally performed. In response to this RAI, Updated Final Safety Analysis Report (UFSAR) Sections B.6.4.3 and B.11.2.1 have been modified to more fully address occupational exposure. New UFSAR Tables B.11-1 and B.11-2 are added.

The exposure estimate for OS197 TC operations is 2.3 person-rem, slightly lower than the 2.5 person-rem exposure estimate for the EOS-TC125 with the EOS-89BTH DSC.

Impact:

UFSAR Sections B.6.4.3 and B.11.2.1 have been revised as described in the response. UFSAR Tables B.11-1 and B.11-2 have been added as described in the response.

Page 41 of 41

ML20190A138

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