ML22115A198
| ML22115A198 | |
| Person / Time | |
|---|---|
| Site: | 07109392 |
| Issue date: | 04/28/2022 |
| From: | Storage and Transportation Licensing Branch |
| To: | NAC International |
| P SAVEROT NRC/NMSS/DFM/STLB 3014157505 | |
| Shared Package | |
| ML22115A196 | List: |
| References | |
| EPID L-2022-NEW-0000 | |
| Download: ML22115A198 (13) | |
Text
Request for Supplemental Information and Observations for the Model No. OPTIMUS-H Package Docket No. 71-9392
This request for supplemental information (RSI) identifies information needed by the staff in connection with its acceptance review of the NAC International (NAC) Model No. OPTIMUS-H package.
CHAPTER 1 GENERAL INFORMATION
1-1 Provide tolerances for all Important to Safety (ITS) components on the licensing drawings so that the as-built package can be verified as being manufactured in accordance with the analyzed design.
All tolerances that support analyses for all ITS components in the application shall be placed on the licensing drawings so that the staff can assess whether the results of the analyses including structural, shielding, and thermal analyses are valid. Tolerances for non-ITS components shall also be included, if they have a potential to impact any ITS component performance.
This information is required to determine compliance with Title 10 of the Code of Federal Regulations (10 CFR) 71.33(a)(5), 71.71(c)(7), 71.73(c)(1), and 71.73(c)(3).
Observations
1-2 Provide the classification of important to safety (ITS) components according to categories A, B, and C from NUREG-6407 on the licensing drawings.
For example, all of the containment boundary components should be ITS Category A.
This information is required to determine compliance with 10 CFR 71.101 and 71.107(a).
CHAPTER 2 STRUCTURAL AND MATERIALS EVALUATION
2-1 Justify the use of a graded ASME Code based criteria for the structural design of the OPTIMUS-H cask components. Explain how the choice of design criteria supports the functional requirements of the cask components.
The applicant did not provide an explanation to ensure that the selection of the acceptance criteria for the package components supports their performance targets.
This information is required to determine compliance with 10 CFR71.31(c).
2-2 Provide a summary table listing all the allowable stresses used in the design of a component where code based allowable stresses for specific conditions are computed
Enclosure 1
using code formulation, nomographs or tables. List the applicable condition, the code reference section, and if the values consider staff positions, cite the reference to the document containing the staff position.
The applicant did not provide enough information to ensure that the allowable stresses are based on appropriate sources including any limitations of use.
This information is required to determine compliance with 10 CFR 71.31(c).
2-3 Provide summary tables identifying the stresses at the critical sections for each load case considered in the combinations in reference to Tables 2.1-1 and 2.1-2 showing the loading combinations, under NCT and HAC, used in the OPTIMUS-H cask component design.
The applicant did not provide information needed by staff to ensure the proper selection of the critical sections for each load case.
This information is required to determine compliance with 10 CFR 71.41(a)
2-4 Explain the methodology used in selecting the design stress values for the load combinations and justify how such a selection is representative of the stress experienced by the component at the different critical sections. Relate this information to the locations selected in Figure 2.1-1 identified for design evaluation.
This information is needed to provide assurance that the design values used in the combinations represent the stress conditions under a given loading.
This information is required to determine compliance with 10 CFR 71.41(a).
2-5 Provide a summary table listing the calculated stress in the attachment including connections to the cask body, for each failure mode considered in the design, along with their code allowable stress limits (i) for all lifting attachments, and (ii) for the tie down gussets/brackets.
Use the computed values to tabulate the design margins for each component of the lifting device.
The applicant has not clearly identified the state of stress for each failure mode of the components.
This information is required to determine compliance with 10 CFR 71.45(a) and 45(b).
2-6 Provide a summary table of results of computed differential thermal expansions between the OSV and the CCV. For each temperature considered, provide the thermal expansion/contraction of the component used in computing the differential expansion.
Provide the same level of information for the expansion between the SIA and CCV to demonstrate that there is adequate clearance between the components to accommodate the effects of thermal changes.
The applicant did not provide enough information to evaluate that there is an available space for the thermal expansion or contraction of package components.
This information is required to determine compliance with 10 CFR 71.71(c)(1)
2-7 Provide plots of the stress contours of some intermediate temperature stages of the model shown in Figure 2.6-2 of the application. Consider temperatures within the CCV temperature gradient along with the corresponding OSV thermal gradient.
The staff needs to evaluate the thermal stress change progress through the cross section.
This information is required to determine compliance with 10 CFR 71.43(f) and 71.51(a)(1).
2-8 Provide a plot of the stress contours for internal pressure, along with a plot of the combined stress resulting from both thermal and pressure.
The staff needs to evaluate the pressure stress gradient across the cross section.
This information is required to determine compliance with 10 CFR 71.43(f) and 71.51(a)(1).
2-9 Starting with the initial bolt pretension, provide a summary of the change (consider a few intermediate stages in temperature with the corresponding pressure) in the bolt pretension due to pressure and temperature respectively.
The staff needs to evaluate the change in the rate of bolt pretension with increase in temperature and pressure.
This information is required to determine compliance with 10 CFR 71.71(c)(4).
2-10 Relate the results of bolt pretension to the nominal compression demand of the O-ring.
The staff needs to ensure that the O-ring can maintain its intended function as a confinement barrier.
This information is required to determine compliance with 10 CFR 71.43(f) and 71.51(a)(1).
2-11 Summarize the stress in the CCV because of reduced pressure. Relate these computed reduced pressure stresses to those in the combined load stress tables and the critical section locations.
The staff needs to evaluate the changes in the stress across the critical section from the change in pressure.
This information is required to determine compliance with 10 CFR 71.71(c)(3).
2-12 Provide a basis why the stress increase on the CCV due to increased external pressure is consider negligible and why there is not impact on the SIA.
The staff needs to establish that the SIA function is not impacted by the change in pressure.
This information is required to determine compliance with 10 CFR 71.43(f)(1) and 71.71(c)(4).
2-13 Summarize, for all impact and drop analyses, the maximum stress computed using the stress resulting from each load consider in the load combinations. Identify the locations of maximum stress in relation to the critical sections, as well as the allowable stress to which the combination stress is compared.
The staff needs to ensure that the critical sections remain critical for the drop analysis.
This information is required to determine compliance with 10 CFR 71.71 and 71.73.
2-14 Summarize the buckling analysis tabulating the resulting stresses from the different load drop conditions analyzed. Identify the locations of the worst buckling stress under each drop condition.
The staff needs to determine that the worst bucking loads are considered.
This information is required to determine compliance with 10 CFR 71.71 and 71.61.
2-15 Present, for all impact and drop analyses, the anticipated non-linear behavior. Describe the types of non-linearities that are anticipated and the numerical algorithm used in the finite element definition to capture the anticipated non-linearities.
The staff needs to confirm that the anticipated deformations are captured by the analytical model.
This information is required to determine compliance with 10 CFR 71.71 and 71.73.
2-16 Explain how the material model used in the analysis addresses the material property variations over the range of temperature to which the cask is required to be exposed during a drop or impact during transportation.
This information is needed to support the geometric response of the model in the analysis.
This information is required to determine compliance with 10 CFR 71.71 and 71.73.
2-17 Provide, for each drop analysis (hot and cold), the stresses at the critical sections identified in Figure 2.1-1 along with the displacement with respect to the vertical axis of the cask as Y and radial axis as X.
The staff needs to be able to track the deformed shape of the cask during a drop analysis.
This information is required to determine compliance with 10 CFR 71.71 and 71.73.
2-18 List the rigid-body accelerations experienced by the package for each load drop condition analyzed. Compare these accelerations with those recorded in the Sandia Laboratory tests.
The staff needs this information to confirm that vibrations during transportation are enveloped by the vibrations of the drop conditions.
This information is required to determine compliance with 10 CFR 71.71 and 71.73.
2-19 Provide a time history plot of the displacement at several points of the impact limiter to capture the cross-sectional extent of the foam deformation for all the drop cases analyzed with impact limiters.
This information is needed to support the deformation profile of the impact limiters.
This information is required to determine compliance with 10 CFR 71.71 and 71.73.
2-20 Provide, in Table 2.7-3 of the application, both Pm and Pb individually and then combined for comparison to allowable. Provide a similar Table for the bottom drop.
This information is needed to demonstrate which of the two stresses controls a specific load case.
This information is required to determine compliance with 10 CFR 71.71 and 71.73.
2-21 Discuss the finite element modeling of the impact limiters including the algorithms used in the finite elements to capture the capture the non-linear behavior of the foam.
Discuss how the change in the material property of the foam is captured as the foam deforms.
This information is required to determine compliance with 10 CFR 71.71 and 71.73.
2-22 Plot stress contours around the points of impact and tabulate the values to show maximum stress and the transition to no displacement for all puncture analysis.
The staff needs to establish the extent of the deformation around the impact zone.
This information is required to determine compliance with 10 CFR 71.71 and 71.73.
2-23 Confirm, for analytical models, model consistency by comparing component weights and centers of gravity computed from the model with the Table 2.1-8 physical values of package components included in the analytical model.
The staff needs to verify the consistency between the analytical model and the physical cask.
This information is required to determine compliance with 10 CFR 71.71 and 71.73.
2-24 Provide a comparison of the OPTIMUS-H package with the NAC-UMS package to establish the inertial similarities of the package. Address the weights and centers of gravities of the two packages along with the distribution of mass around the package
center of gravity. Establish the centroid of the packages and compare the eccentricity between the center of gravity and centroid.
The staff needs to verify the consistency between the analytical model and the physical cask.
This information is required to determine compliance with 10 CFR 71.41(a).
2-25 Compare the results of the natural frequency of elements determined in Section 2.12.4 of the application to those that are derived using the analytical model.
The staff needs to verify that the model frequencies are representative of the frequencies considered in the equivalent static analysis.
This information is required to determine compliance with 10 CFR 71.71 and 71.73.
2-26 Explain why the validation of LS-DYNA was discussed instead of using LS-DYNA to simulate the test drop and compare the results of testing to those using the analytical model.
The applicant did not justify the approach taken in lieu of a more direct approach.
This information is required to determine compliance with 10 CFR 71.73.
CHAPTER 3 THERMAL EVALUATION
3-1 Provide the results from the conservative bounding damage configuration for the bounding case to verify the results from the realistic damage configuration for the base case.
The applicant stated, in SAR section 3.4.3.1, that two different package configurations are modeled to evaluate the thermal response of the packaging during the HAC thermal test with cumulative damage resulting from the HAC top end drop and HAC top end puncture: a realistic damage configuration for the base case and a conservative bounding damage configuration for the bounding case. The temperatures of package components from the realistic damage configuration for the base case are summarized in Table 3.1-4. However, the temperatures of the package components from the conservative bounding damage configuration for the bounding case are not provided in the application.
This information is required to determine compliance with 10 CFR 71.73(c)(4).
3-2 Clarify the use of the radiolysis gas term ratios of (n2/n1 = 1.105) for Aerosol Cans, Type 2 (Content 1-2B) and (n2/n1 = 1.075) for Standard DOT 3E Lecture Bottles (Content 1-2C) in the pressure calculations.
The applicant stated, in section 4.5.5.1.2, Aerosol Cans, Type 2 (Content 1-2B), that the total pressure in the CCV cavity is calculated including the addition of the liquefied gas propellant from a given number rupturing cans, and based radiolysis gas term n2/n1=1.12 for a gas mixture of 8 vol% hydrogen and 4 vol% oxygen, resulting in one the following pressure equation and parameters. However, instead of n2/n1 = 1.12, the
ratios of n2/n1=1.105 (7 vol% hydrogen and 3.5 vol% oxygen) and n2/n1=1.075 (5 vol%
hydrogen and 2.5 vol% oxygen) are used for Aerosol Cans, Type 2 (Content 1-2B) and Standard DOT 3E Lecture Bottles (Content 1-2C), respectively, in the pressure calculations.
The applicant may need to perform pressure calculations with n2/n1=1.12 and update the maximum NCT and HAC pressures of Content 1-2B and Content-1C in SAR Table 4.5-5.
This information is required to determine compliance with 10 CFR 71.71 and 71.73(c)(4).
3-3 Provide a verification and demonstration for items (a) and (b), described below.
The applicant stated in SAR section 3.5.2.1 that reliable calculational models have not been yielded to accurately determine thermal resistance between contacting components. Instead, a subjective approach, as described in SAR section 3.3, is taken to assign the level of thermal resistance at each of the modeled interfaces between contacting components. In the approach, the mixed TCC values of 1,000, 15, 5.0, 1.0 and 0.5 Btu/hour-in2-°F are assigned to contacting component interfaces, respectively with low thermal contact resistance, low/moderate thermal contact resistance, moderate thermal contact resistance, high/moderate thermal contact resistance, and high thermal contact resistance in thermal analyses. These TCC values are obtained experimentally and are determined by the materials of the mating surfaces, the interstitial fluid, surface finish of the mating materials, and the pressure applied to the mating materials. The applicant performed a sensitivity analysis and concluded that TTC values used in the NCT thermal analysis do not change the packaging temperatures by more than 1°F.
Provide the information that demonstrates that (a) the TCC values, assigned to the modeled contact elements, exactly represent the thermal contact conditions between the components based on configuration of the OPTIMUS-H package with a heat load up to 1,500 watts, and (b) the conclusion for the NCT sensitivity analysis ( 1°F) is also applicable to HAC in which the fire temperature is up to 1475°F for a time period of 30 minutes.
Also, the applicant may demonstrate conservatism by assigning the perfect contact (TCC = 1000 Btu/hr-in2-°F) to the model contact elements in the HAC 30-minute fire and the high contact resistance (TCC = 0.5 Btu/hr-in2-°F) to the model contact elements in the post-fire cooldown.
This information is required to determine compliance with 10 CFR 71.73(c)(4).
OBSERVATION
3-4 Clarify typos or inconsistencies for Questions (a) through (e) below in the application.
(a) Section 3.3.1: Revise typos of 114°F (45 °F °C) for ILW contents TRU waste and 177°F (80°F °C) for fuel waste contents as shown in SAR page 3.3-26.
(b) Section 3.3.2: For TRU waste with a thermal load of 200 W, the corresponding average CCV inner surface temperature is 173°F, as shown in Table 3.3-8.
However, Table 3.3-8 shows the average CCV inner surface temperature is 161°F.
(c) Section 3.4.3.1:
(1) Add words (underlined) to the text in SAR page 3.4-6 for clarification: Per the requirements of 10 CFR 71.73(c)(4), the package with TRU waste or irradiated fuel waste (IFW) is evaluated for hypothetical accident conditions.
(2) Page 3.4-7: As shown in Figure 3.4-8, the CCV seal reaches a maximum temperature of 251°F (122°C) and the CCV fill gas reaches a maximum average temperature of 328°F (164°C) with a content heat load of 200 W. However, Figure 3.4-8 shows a maximum temperature of 265.1°F (129.5°C) for the CCV seal and a maximum average temperature of 341°F (172°C) for the CCV fill gas.
(3) Page 3.4-7: As shown in Figure 3.4-9, the CCV seal reaches a maximum temperature of 335°F (168°C) and the CCV fill gas reaches a maximum average temperature of 635°F (335°C) with a content heat load of 1,500 W. However, the staff finds Figure 3.4-9 shows maximum temperatures of 339°F (170.6°C) at the contact side and 350.8°F (177.1°C) at opposite side for the CCV seal and a maximum average temperature of 646.9°F (341.6°C) for the CCV fill gas.
(4) Page 3.4-8: The maximum average temperature of the CCV lid at the bolted closure reaches its maximum of 328°F (164°C) at 7.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> into the transient HAC analysis and the temperature contours within the CCV lid at this time are presented in Figure 3.4-12. However, Figure 3.4-12 shows a range of 330.7°F~350.8°F indicating the maximum average temperature of the CCV lid should be higher than 328°F.
(d) Table 3.4-3: For both HAC3 and HAC4 under Fuel Waste, revise (200W, He) to (1,500W, He).
(e) Section 3.1.3: As shown in Table 3.3-8, the maximum temperature of the accessible surface of a package with ILW contents TRU waste when exposed to an ambient temperature of 100°F in still air and shade is 114°F (45°C), below the 122°F (50°C) temperature limit for nonexclusive use.
This information is required to determine compliance with 10 CFR 71.71 and 71.73(c)(4).
CHAPTER 5 SHIELDING EVALUATION
5-1 Provide specifications for the contents to be shipped and justification for the assumed source terms in the contents.
Page 1.1-1 of the application states: The package radioactive contents include Type B quantities of normal form Transuranic (TRU) waste and Irradiated Fuel Waste (IFW) contents. On page 5.1-3 of the application, it is further stated: Because the contents of the package are variable, and the final isotope inventory is different for each TRU waste content or irradiated fuel waste liner, the maximum dose rates are strongly dependent on the specific contents. Additionally, as the package may be shipped as open or closed transport, the limiting dose rates are variable as well.
As an example, maximum dose rates are calculated for the maximum allowable quantity of two individual isotopes. The two isotopes considered are Co-60 (), in Tables 5.1-2 (NCT) and Table 5.1-3 (HAC) and Cf-252 (n and ) in Table 5.1-4 (NCT) and Table 5.1-5 (HAC). The application further states: TRU waste is contaminated tools and materials
used in processing facilities and irradiated fuel waste is distributed through a fuel volume.
In addition, page 7-1 of the application states: All contents to be shipped shall satisfy the requirements for type and form of material, maximum quantity of contents per package, and loading restrictions described in Section 1.2.2. However, the information provided in Section 1.2.2 is insufficient for the staff to assess whether the sources from these two isotopes will bound all contents to be shipped with this packaging system because TRU waste is typically very complex in radioactive material composition and hence the source terms. For example, U-232 in the TRU wastes will decay to Tl-208 which emits a 2.6-MeV gamma and accounts for about 85 percent of the total dose from U-232.
This information is required to determine compliance with 10 CFR 71.47 and 71.51.
5-2 Provide shielding calculations for the package configuration that uses the one-inch shield insert assembly (SIA) without a closure lid.
The packaging system includes the use of three SIA designs. The three SIAs are provided in 1-inch, 21/4-inch and 33/4-inch thicknesses. Page 1.1-1 of the application states: The SIA is a painted carbon steel container consisting of a body and lid fully encasing the radioactive contents to provide additional shielding. The 1-inch SIA does not include a lid. However, the application does not include the dose rates for the package that uses the 1-inch SIA.
This information is required to determine compliance with 10 CFR 71.43, 71.47, and 71.51.
5-3 Provide information on how the uncertainties of MCNP shielding calculations are treated in the final dose rates and revise the calculation results as necessary.
Page 5.4-5 of the application states: This, along with a check of the reported fractional standard deviation () for each tally bin and the additional statistical information reported for MCNP tallies, ensures the reliability of all MCNP calculated dose rate results. Page 5.4-5 further states: Table 5.4-2 and Table 5.4-3 list the calculated NCT dose rates for each energy group for photons and neutrons, respectively. The results in these tables are based directly on the results from MCNP and include the 2 uncertainty.
However, the MCNP users manual indicates that the statistical uncertainty given by the code for shielding calculation is the relative error (r) to the mean value rather than a standard deviation. For problems with large mean values and relatively large relative errors, the final results between adding the standard deviation to the mean value (x +
- 2) and adding the mean value multiplied by the relative error ((1 + r)
- x) can be significantly different (x is the mean value, is the standard deviation, and r is the relative error).
This information is required to determine compliance with 10 CFR 71.47 and 71.51.
Observation 5-1
Justify the SS-304 density as being conservative for shielding calculation or revise the shielding calculation with appropriate material properties for the SS-304.
In Table 5.3-3 of the application, the applicant provided the material property used in the shielding calculations for the package. However, the staff notes that the density of the SS-304 stainless steel is given as 8.0 g/cc. This value is larger than the 7.82 gg/cc value recommended for shielding analyses as published in PNNL-15870 Rev. 1, Compendium of Material Composition Data for Radiation Transport Modeling, and the standard material composition provided in SCALE code. Because the application assumed a non-conservative material density in the shielding calculations for all SS-304 components, the calculated dose rates could be non-conservative. As such, it may be necessary to redo the shielding calculation using appropriate material properties for the SS-304 components.
This information is required to determine compliance with 10 CFR 71.47 and 71.51.
CHAPTER 6 CRITICALITY
6-1 Demonstrate that the skin of the impact limiter will be able to support the dead weight of the canister and contents under hypothetical accident conditions so that the credit for top impact limiter spacing of 35% is appropriate for the criticality safety analysis.
Page 6.3-1 of the application states: The impact limiters are comprised of a 1/2-inch thick SS inner shell and a 14-gauge stainless steel outer skin filled with polyurethane foam.
The application further states: For the HAC package array, a top impact limiter spacing of 35% (3.53 in, 8.9611 cm), with no impact limiter material modeled, is credited for the top and bottom of the package in the HAC package arrays. Similarly, a side impact limiter spacing of 29% (3.53 in, 8.9611 cm), with no impact limiter material modeled, is credited for the side of the package in the HAC package arrays. However, it is not clear if the skin of the 14-gauge stainless steel (0.0747) skin of the impact limiter under HAC will be able to support the weight of the package so that the credit can be taken in criticality calculations.
Similarly, it is not clear either if the 3.53-inch side spacing is appropriate for the package under HAC, i.e., first with a nine-meter drop and then an 800°C fire for 30 minutes.
This information is required to determine compliance with 10 CFR 71.59.
6-2 Clarify whether the contents are TRU wastes or weapon grade/aged weapon grade plutonium and perform criticality safety separately for each group of the contents.
The applicant provided the material composition for the material composition in Table 6.3-4. Note c to the table states: Density is constant for Materials 238-244 and represents: Plutonium, Aged WGPu (A: 4-7% Pu-240); Plutonium, Aged WGPu (B: 10-13% Pu-240); Plutonium, Aged WGPu (C: 16-19% Pu-240); Plutonium, DOE 3013 WGPu; Plutonium, Fuel Grade; Plutonium, Power Grade; and Plutonium, Shefelbine WGPu. This note appears to indicate that the contents could be TRU wastes, weapon grade or aged weapon grade plutonium. If so, separate criticality safety analyses are needed for each content type.
This information is required to determine compliance with 10 CFR 7.55 and 71.59.
Observation
6-3 Provide means for controlling the size of the fissile materials in the contents.
The applicant performed various studies on the dependence of the neutron multiplication factor keff and the particle size. The results show that there is a strong dependence between the keff of the package and particle size. For example, Table 6.6-8, FEM-1, Sphere Particle Baseline Configuration - HAC Package Array shows that the keff of the package increases as the particle size increases and peaks when the particle size reaches 2 cm in diameter.
Table 6.6-8 also shows that the keff of the package will be greater than the upper subcritical limit (USL) when the particle diameter is in the range of 0.5 cm to 4 cm. While the results are in general consistent with the basic theory of fission system [Ref. 1], it is not clear how the particle size in the TRU wastes is controlled so that keff of the package remains below the USL.
This information is required to determine compliance with 10 CFR 71.55 and 71.59.
Reference
- 1. J. J. Duderstadt and L. J. Hamilton, Nuclear Reactor Analysis, page 405, JOHN WILEY & SONS, Inc, New York, 1976.
CHAPTER 7 OPERATING PROCEDURES
7-1 Provide a method for the users of this package design to determine the allowable contents.
10 CFR 71.5(a) requires Each licensee who transports licensed material outside the site of usage, as specified in the U.S. Nuclear Regulatory Commission license, or where transport is on public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the DOT regulations in 49 CFR parts 107, 171 through 180, and 390 through 397, appropriate to the mode of transport. 49 CFR 173.403 states: The consignor must provide to the initial carrier specific written instructions for maintenance of exclusive use shipment controls, including the vehicle survey requirement of § 173.443(c) as applicable, and include these instructions with the shipping paper information provided to the carrier by the consignor.
Therefore, a clear instruction for determining the allowable contents must be included in the operating procedures of the package. Because this package design relies on different methods for determining the allowable contents for meeting the different parts, i.e., shielding, criticality, and thermal, of the regulatory requirements, it is imperative to include instructions for determining the allowable contents as part of the certificate by references. The instructions shall be established based on the intersections of the limits for shielding, criticality, and thermal safety analyses.
This information is required to determine compliance with 10 CFR 71.5.
7-2 Provide information on items (a), (b) and (c) below in the application (e.g., SAR Chapter
- 7) for review.
The applicant presented the cask containment vessel (CCV) free volume and fill gas temperatures in SAR Table 7.5-9 for various assembly configurations of the bare basket,
1-inch Shield Insert Assembly (SIA)-98, 2.25-inch SIA-97 and 3.75-inch SIA-96 and the CCV TRU waste cavity gas temperature, as a function of content decay heat, in SAR Figure 7.5-2.
In addition to the information provided in Table 7.5-9 and Figure 7.5-2 in SAR Chapter 7, the applicant needs to provide information on the items below for review:
(a) descriptions regarding Table 7.5-9 and Figure 7.5-2 (b) calculations of the CCV free volume for all assembly configurations listed in Table 7.5-9, and (c) derivations (or references) of the polynomial equations y=f(x), the NCT fill air/helium temperature vs. decay heat, as shown in Figure 7.5-2.
This information is required to determine compliance with 10 CFR 71.71 and 71.87.
Observations
7-3 Describe how the loading of the non-compliant TRU waste will meet the allowable pressure limit and the design heat load during the loading operations.
The applicant stated, in SAR section 3.3.2, that the pressures for the contents with non-compliant pressurized items are much higher because the allowable quantities of non-compliant items for each waste type is determined based on the maximum quantity that still results in a final pressure under the NCT design pressure of 100 psig, as shown in SAR Table 3.3-10.
With the contents optimized based on the final allowable pressure and the small pressure margins below the NCT design pressure of 100 psig, the applicant may need to describe, in SAR Chapter 7, how the loading of the non-compliant TRU waste will meet the limits of pressure and decay heat during the loading operations.
This information is required to determine compliance with 10 CFR 71.87.
ML22115A196; ML22115A198 OFFICE NMSS/DFM/STLB NMSS/DFM/NARAB NMSS/DFM/MSB NMSS/DFM/CTCFB NAME PSaverot PS ZLi ZL TBoyce TBFChang FC DATE Apr 26, 2022 Apr 26, 2022 Apr 27, 2022 Apr 26, 2022 OFFICE NRR/DEX/ESEB NMSS/DFM NMSS/DFM/STLB NMSS/DFM/STLB NAME SSamaddar SS DMarcano DM SFigueroa SFYDiaz-Sanabria YD DATE Apr 26, 2022 Apr 26, 2022 Apr 28, 2022 Apr 28, 2022 OFFICE NMSS/DFM/STLB NAME PSaverot PS DATE Apr 28, 2022