ML23069A287

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Amendment 18 to Standardized NUHOMS Certificate of Compliance No. 1004 for Spent Fuel Storage Casks Non-Proprietary Safety Analysis Report Page Changes
ML23069A287
Person / Time
Site: 07201004
Issue date: 01/20/2023
From:
Orano TN Americas
To:
Office of Nuclear Material Safety and Safeguards
Shared Package
ML23020A920 List:
References
E-61864
Download: ML23069A287 (1)
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Text

Enclosure 6 to E-61864

Proposed Amendment 18, Revision 2 Changes to the Standardized NUHOMS System Updated Final Safety

Analysis Report (Public)

3.2 Structural and Mechanical Safety Criteria The reinforced concrete HSM and its DSC support structure, the DSC and its internal basket assembly, and the transfer cask are the NUHOMS system components which are important to safety. Consequently, they are designed and analyzed to perform their intended functions under the extreme environmental and natural phenomena specified in 10CFR72.122 (3.6) and ANSI-57.9 (3.36). Since the NUHOMS ISFSI is an independent, passive system, no other components or systems contribute to its safe operation.

Table 3.2-1 summarizes the design criteria for the standardized NUHOMS system components which are important to safety. This table also summarizes the applicable codes and standards utilized for design. The extreme environmental and natural phenomena design criteria discussed below comply with the requirements of 10CFR72.122 and ANSI-57.9. A description of the structural and mechanical safety criteria for the other design loadings listed in Table 3.2-1, such as thermal loads and cask drop loads, are provided in Chapter 8. The principal design criteria for the NUHOMS61BT system are provided in Appendix K.

The principal design criteria for the NUHOMS HSM Model 80 and Model 102 described in this chapter are also applicable to HSM Model 152 and HSM Model 202. See Appendi x R and Appendix V, respectively, for details. See Appendix P for HSM-H design criteria and Appendix U for the high seismic design criteria for the HSM-HS.

3.2.1 Tornado and Wind Loadings The NUHOMS ISFSI is designed to be located anywhere within the contiguous United States. Consequently, the most severe tornado and wind loadings specified by NRC Regulatory Guide l.76 (3.7) and NUREG-0800, Section 3.5.1.4 (3.8) are selected as the design basis. The NUHOMS reinforced concrete HSMs are designed to safely withstand 10CFR72.122 (b)(2) tornado missiles. Extreme wind effects are much less severe than tornado wind and missile loads or seismic effects and, therefore, are not evaluated in detail for the HSM.

Since the NUHOMS on-site transfer cask is used infrequently and for short durations, the possibility of a tornado funnel cloud enveloping the cask/DSC during transit to the HSM is a low probability event. Nevertheless, the transfer cask is designed for the effects of tornados, in accordance with 10CFR72.122. This includes design for the effects of worst case tornado winds and missiles.

For short term operations which are not analyzed for tornado hazards, administrative controls during ISFSI handling operations are required. Those controls, which are described in Section 5.1.1.5, coupled with the low probability of tornado events, gives confidence in weather conditions being acceptable during outdoor operations during the short durations when the system configuration was not analyzed for tornado hazards.

NOTE: Though the administrative controls are detailed in Section 5.1.1.5 on transfer cask downending and transfer, they apply to any short-term operations that are not analyzed for tornado hazards, potentially including, for example, infrequently performed maintenance and inspection related activities, and performance of aging management activities.

January 2023 Revision 2 72-1004 Amendment No. 18 Page 3.2-1 During DSC insertion/withdrawal operations, the transfer cask is docked with the HSM docking flange and mechanically secured to embedments provided in the front wall of the HSM. The cask restraints used for this purpose are shown in Figure 4.2-13. The embedments are equally spaced on either side of the HSM access opening. The HSM embedments are designed in accordance with the requirements of ACI 349 Code (4.14).

The transfer cask restraint system is designed for loads which occur during normal DSC transfer operations and during an off-normal jammed DSC event.

The HSM gap between modules is covered with stainless steel wire bird screen to prevent pests or foreign material from entering the HSM. Periodic surveillance constitutes the only required maintenance activity for the NUHOMS ISFSI.

It is expected that during the installation and loading of an HSM array there will be empty modules. Vacant HSMs can occur due to: partial filling of a complete construction phase of HSMs, or a partial filling of a phase of HSMs which will be expanded at a future date. The following issues have been evaluated for both cases:

Normal Operation Issues, Construction Issues, and Accident Condition Issues. During installation of an additional HSM(s), or for other reasons, shield wall(s) may be removed for a period of time. However, compensatory measures shall be considered for radiation shielding and for missile protection, i f necessary.

The design flexibility of the HSMs permits a licensee to choose the most economical arrangement of HSMs which best meets plant specific conditions and requirements. This SAR presents a detailed analysis for a single stand-alone module as this is the governing design case for the postulated environmental loads such as earthquake, flooding, and tornado loads. Thermal loads also provide significant loadings for the HSM structural design for the free-standing prefabricated HSM.

A typical reinforcing steel layout for the HSM floor, walls, and roof is shown in Figure 8.1-19. The reinforcement sizing and placement specified is used for HSM array configurations ranging in size from a single stand alone module to a 2x10 array of HSMs or larger. Licensing details, such as concrete joint and reinforcing bar lap splice requirements, are shown on the Appendix E drawings.

The HSM design documented in this SAR is constructed of 5,000 psi (minimum) compressive strength, normal weight (145 pounds per cubic foot minimum density) concrete with Type II Portland cement meeting the requirements of ASTM C150 (4.6) or blended Portland cement meeting the requirements of ASTM C595 [4.8]. Type III cement may be used as long as it meets the chemical and physical requirements for Type II cement specified in ASTM C150. This applies to all HSM models including HSM-H and HSM-HS unless testing is performed. The concrete aggregate meets the specifications of ASTM C33 (4.6). The concrete is reinforced by ASTM A615 or A706 Grade 60 (4.7) deformed bars placed vertically and horizontally at each face of the walls, roof and floor.

January 2023 Revision 2 72-1004 Amendment No. 18 Page 4.2-7 4.10 References

4.1 U.S. Government, Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation (ISFSI), Title 10 Code of Federal Regulations, Part 72, Office of the Federal Register, Washington, D.C.

4.2 Deleted.

4.3 Deleted.

4.4 Deleted.

4.5 American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code,Section III, Division 1, 1983 Edition, with Winter 1985 Addenda.

4.6 American Society for Testing and Materials, Annu al Book of ASTM Standards, Section 4, Volume 04.02, 1990.

4.7 American Society for Testing and Materials, Annual Book of ASTM Standards, Section 1, Volume 01.04, 1990.

4.8 ASTM International, C595/C595M-21, Standard Specification for Blended Hydraulic Cements.

4.9 American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More for Nuclear Materials, ANSI N14.6 -

1993, American National Standards Institute, Inc., New York, New York.

4.10 American Concrete Institute, Building Code Requirement for Reinforced Concrete, ACI-318, 1983.

4.11 American Institute of Steel Construction, (AISC), Specification for Structural Steel Buildings, Ninth Edition 1989, Chicago, Illinois.

4.12 American National Standards Institute, American National Standard for Radioactive Materials - Leakage Tests on Packages for Shipment, ANSI N14.5, 1977.

January 2023 Revision 2 72-1004 Amendment No. 18 Page 4.10-1

28. Open the valve on the vent por t and allow helium to flow into the DSC cavity to pressurize the DSC to 2.5 psig +/- 2.5 psig in accordance with Technical Specification 3.1.2.a limits.
29. Close the valves on the helium source.
30. Remove the Strongback, decontaminate as necessary, and store.

5.1.1.4 DSC Sealing Operations

1. Disconnect the VDS from the DSC. Seal weld the prefabricated plugs over the vent and siphon ports and perform a dye penetrant weld examination in accordance with the CoC Appendix A Inspections, Tests, and Evaluations Item 4.3 requirements.
2. Install the automated welding machine onto the outer top cover plate and place the outer top cover plate with the automated welding system onto the DSC. Verify proper fit up of the outer top cover plate with the DSC shell.
3. Tack weld the outer top cover plate to the DSC shell. Complete the outer top cover plate weld root pass. Perform dye penetrant examination of the root pass weld.

Weld out the outer top cover plate to the DSC shell and perform dye penetrant examination on the we ld surface in accordance with the CoC Appendix A Inspections, Tests, and Evaluations Item 4.3.

4. Remove the automated welding machine from the DSC. Rig the cask top cover plate and lower the cover plate onto the transfer cask.
5. Bolt the cask cover plate into place, tightening the bolts to the required torque in a star pattern. Verify that the TC radial dose rates measured at the surface of the transfer cask are compliant with limits specified in CoC Appendix A Inspections, Tests, and Evaluations Item 3.2. The configuration for determining the TC radial surface dose rates shall be in accordance with CoC Appendix A Inspections, Tests, and Evaluations Item 3.2.

5.1.1.5 Transfer Cask Downending and Transfer to ISFSI

For short term operations which are not analyz ed for tornado hazards, the following administrative controls during ISFSI handling operations are required.

NOTE:

Although these administrative controls are detailed in this section on transfer cask downending and transfer, they apply to any short-term operations which are not analyzed for tornado hazards, potentially including, for example, infrequently performed maintenance and inspection related activities, and performance of aging management activities.

General Licensees will develop, revise, or review existing procedures to establish administrative controls that implement compensatory measures to mitigate tornado

January 2023 Revision 2 72-1004 Amendment No. 18 Page 5.1-8a hazards during periods of ISFSI handling operations that include the following considerations:

Prior to starting ISFSI handling operations:

1. Licensees should consider performing site walkdowns to identify and secure any potential hazards that could hamper short-term operations during periods of adverse weather or during periods when adverse weather is predicted.
2. The licensee should determine the expected bounding duration of the handling operation by either benchmarking or dry runs. The weather checks, at a minimum, should then be for the expected duration of the handling operation, including a margin for contingency. Subsequently, that duration should then be periodically assessed based upon operating experience.
3. In coordination with the Control Room or other ISFSI operational authority, the local weather forecast shall be confirmed acceptable. The weather forecast shall indicate that there are no tornado watches, ad visories, or warnings within the expected duration of the handling operation.
4. Regarding the local weather forecast, licensees may use information from the National Weather Service (NWS) unless another resource for the site (e.g., the National Oceanic At mospheric Admin istration, Weather Forecast Office nearest the site, weather channel, etc.) is already approved for use or can be justified as providing equivalent information in terms of timeliness and accuracy.
5. The licensee should prescribe specific ti mes that the we ather forecast would be checked, the area(s) considered, and the frequency of the forecast checks, accounting for the site configuration and time necessary to be able to bring the system into an analyzed configuration at any time, if necessa ry, when there is a change in forecast.
6. The prescribed timeframe for performing weather checks should be no greater than the expected bounding duration of the handling operation established in Item 2 above thereafter until ISFSI handling operations are completed.
7. A log or checklist shall be established to document weather checks prior to starting handling operations.
8. Satisfactory completion of these criteria should be recorded and maintained with the documentation for the DSC campaign.

During ISFSI handling ope rations:

1. Staff shall be assigned to monitor weather during handling operations.
2. Maintain a log or checklist that documents weather checks during handling operations.

January 2023 Revision 2 72-1004 Amendment No. 18 Page 5.1-8b

3. Any time the above conditions cannot be met during handling operations, the storage system SSCs shall be placed in a safe and analyzed condition as soon as practicable.

The duration of ISFSI handling operations during which ISFSI important to safety SSCs are in an unanalyzed condition shall be minimized to the extent practicable. For NUHOMS systems, those conditions are:

  • From the time a loaded transfer ca sk (TC) lid is removed, when the dry shielded canister is to be inserted into a horizontal storage module (HSM), until the time the HSM door is installed.
  • Any other time a loaded T C lid is removed outdoors, for site configuration requirements or an HSM door is removed.
  • During the time a T C is rotated from a vertical to a horizontal orientation, outdoors, for site configuration requirements.

NOTE:

Alternate Procedure for Downending of Transfer Cask : Some plants have limited floor hatch openings above the cask/trailer/skid, which limit crane travel (within the hatch

January 2023 Revision 2 72-1004 Amendment No. 18 Page 5.1-8c NUHOMS CoC 1004 APPENDIX A INSPECTIONS, TESTS, AND EVALUATIONS AND APPENDIX B TECHNICAL SPECIFICATIONS BASES

As discussed on page 10-1, with Amendment 11 to CoC 1004, the Technical Specifications (TS) were converted to the NUREG-1745 format and the TS bases were returned to this chapter. The numbering scheme for the TS changed a great deal as the TS were converted to the NUREG -

1745 format. Additionally, as also discussed on Page 10-1, Amendment 16 further changed the numbering scheme for the licensing basis documents. Therefore there is not a documented basis for each CoC condition, ITE item, or TS and therefore the numbering scheme of this chapter, as reflected in the table of contents below, is not comprehensive.

TABLE OF CONTENTS

B 10 ITE 3.1 Site Specific Parameters and Analyses......................................................... 10-12 B 10 ITE 3.2 Transfer Cask Dose Rates............................................................................. 10-13 B 10 ITE 3.3 HSM or HSM-H Dose Rate Evaluation Program......................................... 10-14 B 10 ITE 4.1 Leak Test....................................................................................................... 10-15 B 10 ITE 4.3 DSC Dye Penetrant Test of Closure Welds.................................................. 10-15 B 10 TS 2 FUNCTIONAL AND OPERATING LIMITS............................................. 10-18 B 10 TS 3 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY.. 10-25 B 10 TS 3 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY................... 10-28 B 10 TS 3.1 DSC FUEL INTEGRITY............................................................................. 10-32 B 10 TS 3.1.1 DSC Bulkwater Removal Medium and Vacuum Drying Pressure............... 10-32 B 10 TS 3.1.2 DSC Helium Backfill Pressure..................................................................... 10-35 B 10 TS 3.1.3 Time Limit for Completion of Transfer Operations (24PTH, 61BTH Type 2 or 32PTH1, 69BTH or 37PTH DSC Only)..................................... 10-37 B 10 TS 3.2 CASK CRITICALITY CONTROL.............................................................. 10-39 B 10 TS 3.3.1 Maximum DSC Removable Surface Contamination.................................... 10-41 B 10 TS 4.3.2 RADIATION PROTECTION PROGRAM.................................................. 10-42 B 10 TS 4.3.3 Hydrogen Gas Monitoring for Specified DSCs............................................ 10-42 B 10 TS 4.3.6 HSM or HSM-H Thermal Monitoring Program........................................... 10-42 B 10 TS 4.4 CASK TRANSFER CONTROLS................................................................ 10-43 B 10 TS 4.4.1 TC/DSC Lifting/Handling Height Limits..................................................... 10-43 B 10 TS 4.4.2 Supplemental Shielding Drop onto OS197L TC.......................................... 10-45

January 2023 Revision 2 72-1004 Amendment No. 18 Page 10-11 Proprietary and Security Related Information for Drawing NUH24PTH-1002-SAR, Rev. 3B Withheld Pursuant to 10 CFR 2.390 P.4.4.7 HSM-H Thermal Model Results P.4.4.7.1 Normal and Off-normal Operating Condition Results Temperature distributions for the normal and off-normal cases are show n in Figure P.4-6 through Figure P.4-13. The maximum component temperatures for the normal and off-normal cases are listed in Table P.4-2, Table P.4-3, and Table P.4-4. Temperature distributions for the single HSM -H which provides maximum temperature gradients in concrete walls, are shown in Figure P.4-16. Note that Figure P.4-16 shows the analysis temperature distribution before any adjustments made based on the results for bounding Case 1 documented in Table P.4-2. As seen from Table P.4-2 and Table P.4-3, the HSM-H concrete and DSC shell temperatures without the fins on the side heat shield for 31.2 kW are bounded by the case with the fins for 40.8 kW decay heat load. Therefore, fins are not required on the side heat shields in the HSM-H, if the total he at load is 31.2 kW or less. This is summarized in Table P.4-43.

P.4.4.7.2 Accident Condition Results Temperature distributions for the blocked vent accident case with 40.8 kW decay heat load at 38.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after blockage of the vents are shown in Figure P.4 -14. The maximum component temperatures for the blocked vent accident case are listed in Table P.4-5.

Figure P.4-15 shows the time-temperature history of HSM-H components for this transient.

The maximum component temperatures for these cases are listed in Table P.4-6. Figure P.4-17 provides maximum temperature gradients in concrete walls during accident conditions. Table P.4-5 and Table P.4-6 incorporate the adjustments made to the analytical results as described in P.4.4.8 based on the thermal tests of the HSM-H [4.48]. Note that Figure P.4-14, Figure P.4-15 and Figure P.4 -17 show the analysis temperature distributions, before any adjustments made b ased on the results for bounding Case 1 documented in Table P.4-2.

P.4.4.8 Evaluation of HSM-H Performance The thermal performance of the HSM-H is evaluated under normal, off-normal, and accident conditions of operation as described above and is shown to satisfy all the temperature limits and criteria. The DSC shell temperatures calculated here, are used in the DSC basket and fuel cladding models as a boundary condition in Section P.4.6. The results show that all the basket and fuel cladding material temperature limits are satisfied. The results of the HSM-H temperatures are used in Section P.3 toshow that thermal stresses in the HSM-H are also within these allowables.

The results of the 117 °F ambient blocked vent condition show that the maximum concrete temperature at the end of 38.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (with finned side heat shields, louvered top heat shield, and with slots on plate on top of support rail) and 30.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (with flat stainless steel heat shields and without slots on plate on top of support rail) in the blocked vent accident are 431 °F and 415 °F, respectively. These are above the 350

°F limit given in NUREG 1536 [4.42] for a ccident conditions. To account for the effect of higher concrete temperature on the concrete compressive strengths, the structural analysis of HSM-H concrete components in Section P.3 is based on 10% reduction in concrete material properties. Elevated temperature testing of the concrete mix (cement type, additives, water -cement ratio, aggregates, proportions) is performed to ensure that the required strength is maintained. Portland cements meeting the requirements of ASTM C 150 or ASTM C595 (blended Portland cement) are acceptable for use. The use of any Portland cement concretes w ill require testing to be performed when the concrete accident temperature exceeds 350 °F. Testing will be performed to demonstrate that the level of strength reduction is less than the 10% reduction that was emp loyed in the calculations, and to ensure that there is no deterioration of the concrete due to higher temperatures.

January 2023 Revision 2 72-1004 Amendment No. 18 Page P.4-25 P.5.4.6.1 Source Term Assumptions

  • The primary neutron source in LWR spent fuel is the spontaneous fission of 244Cm. For the ranges of exposures, enrichments, and cooling times in the fuel qualification tables, 244Cm represents more than 85% of the total neutron source. The neutron spectrum is, therefore, relatively constant for the fuel parameters addressed herein and is assumed to follow the 244Cm fission spectrum provided in Section P.5.2.2.
  • Surface gamma dose rates are calculated for the HSM and cask surfaces using the actual photon spectrum applicable for each case.
  • The design basis radiological sources are determined with SAS2H\\ORIGEN-S depletion 72.48 models corresponding to 0.490 MTU/FA heavy metal weight.

P.5.4.6.2 HSM-H Dose Rate Analysis Assumptions

  • The 24PTH-L DSC and fuel assemblies are positioned as close to the HSM-H front door as possible to maximize the HSM-H front wall dose rates.
  • Planes of reflection are used to simulate adjacent HSM-Hs.
  • Embedments and rebar in the HSM-H concrete are conservatively neglected.
  • Axial source distribution assumed as shown in Table P.5-13.
  • Fuel is homogenized within the fuel compartment, although the 24PTH-L DSC basket is modeled explicitly.
  • Basket steel plates are modeled as stainless steel. High-strength low-alloy steel (HSLA) is used in the Type 3 basket. The HSLA steel density is 7.83 g/cm3 (or 0.283 lb/in3 see Table P.3.3-10), which is approximately 1% below that of stainless steel. The effect on dose rates is not expected to be notable.

P.5.4.6.3 HSM-Model 102 Dose Rate Analysis Assumptions

  • The HSM-Model 102 is not modeled in MCNP. Dose rates from the HSM-Model 102 analysis from Appendix N.5 are scaled appropriately to account for both the increase in source and decrease in DSC shield plug thicknesses as described in Section P.5.4.7.2.
  • The relative change in the dose rate due to the decreased shield plug thicknesses can be estimated by treating the shield plugs as infinite planes and taking the ratio of attenuation.
  • As it is estimated that the side and roof dose rates presented in Appendix N.5 are highly conservative because the fuel and basket are homogenized, additional scaling factors of 0.67 and 0.8 are introduced for gamma and neutron dose rates, respectively, at the side and roof of the HSM-Model 102 as described in Section P.5.4.7.2.

January 2023 Revision 2 72-1004 Amendment No. 18 Page P.5-19 P.5.4.6.4 OS197FC TC and Standardized TC Dose Rate Analysis Assumptions

  • The 24PTH-L Type 2 and 3 DSCs are modeled within the OS197FC TC. The 24PTH-S DSC is not modeled because it is bounded by the 24PTH-L DSC. The 24PTH-S-LC Type 2 and 3 DSCs are modeled within the Standardized TC. The Type 2 basket bounds the Type 1 basket because the Type 2 basket does not have aluminum inserts in the transition rails. In this chapter, all Type 2 basket results also bound the Type 1 basket. The Type 2 and 3 baskets are explicitly modeled because the basket designs are different. Basket steel plates are modeled as stainless steel.
  • Only the OS197FC is modeled for the welding operation. Three inches of supplemental neutron shielding and one inch of steel are assumed to be placed on top of the 24PTH-L DSC cover plates during welding.
  • During the accident case, the cask neutron shield (either water or NS-3) and the neutron shield jacket (outer steel skin) is assumed to be lost.
  • Axial source distribution assumed as shown in Table P.5-13.
  • Fuel is homogenized within the fuel compartments, although the 24PTH-L DSC and 24PTH-S-LC DSC baskets are modeled explicitly.
  • In the OS197FC TC model, the gap in the cask lid is assumed to extend around the entire circumference of the lid.

P.5.4.7 Normal Condition Models

Three classes of MCNP models are developed: (1) 24PTH-L Type 2 DSC in HSM-H, (2) 24PTH-L Type 2 or 3 DSC in OS197FC TC, and (3) 24PTH-S-LC Type 2 or 3 DSC in Standardized TC. A fourth scenario, 24PTH-S-LC DSC in HSM-Model 102, is analyzed by scaling the results from a similar analysis in Appendix N.5. A fifth scenario, the 24PTH-S-LC DSC inside the OS197/OS197H TC, is shown to be bounded by previously considered scenarios.

These models are described in subsequent sections.

P.5.4.7.1 24PTH-L DSC in HSM-H

Two three-dimensional MCNP4C2 models are developed for the 24PTH-L Type 2 DSC within a HSM-H, one model for neutrons and the other for gammas. These models are presented in Figure P.5-3 through Figure P.5-7. The HSM-H length is designated as the x axis, the width as the y axis, and the height as the z axis. The HSM-H door is designated as the south side and the

-x direction, with the east wall as the -y direction. The roof is the +z direction. The east wall is designated as a reflective boundary and an end shield wall (3 ft thick) is attached to the west wall.

January 2023 Revision 2 72-1004 Amendment No. 18 Page P.5-20 Because the 24PTH-S-LC Type 3 basket design is significantly different than the 24PTH-S-LC Type 2 basket design, an explicit MCNP5 model is developed for the 24PTH-S-LC Type 3 basket based upon the drawings provided in Section P.1.

The 24PTH-S-LC DSC has lead in the top and bottom shield plugs. There are two primary fabrication options for the lead shield plugs. Option A is a poured-lead design and is featured in the original Type 2 basket MCNP models. Option B is a machined-lead design and is featured in the Type 3 basket MCNP models. For option B, 1/8 inch radial gaps are modeled between all lead shield plugs and steel interfaces, and the lead density is reduced from 11.34 g/cm3 to 11.0 g/cm3 (3% reduction in density to account for potential voids in the lead). The shield plug lead thickness is also reduced for option B compared to option A, as indicated in the drawings provided in Section P.1. Therefore, option B results in higher cask end dose rates than option A.

The Standardized TC model, dose rate tallies, and source terms are identical to the Type 2 basket MCNP model. Separate MCNP models are developed for gamma, neutron, and secondary gamma radiation to facilitate the use of weight windows generated by ADVANTG

[5.22]. A cross-sectional view of the Type 3 basket is provided in Figure P.5-25. Dose rates for the 24PTH-S-LC Type 3 DSC in the Standardized TC are provided in Table P.5-5a.

Comparing the dose rates in Table P.5-5 and Table P.5-5a, side dose rates decrease for the Type 3 basket compared to the Type 2 basket due to the different transition rail design. However, because the basket members are generally lighter in the Type 3 basket, the end dose rates increase for the Type 3 basket compared to the Type 2 basket. In addition, end dose rates also increase due to modeling the machined-lead shield plug option.

P.5.4.7.5 24PTH-S-LC DSC in the OS197/OS197H TC

The 24PTH-S-LC DSC may also be transferred in the OS197/OS197H TC. No explicit MCNP models are developed for this configuration because it is bounded by previously analyzed configurations.

Recall the 24PTH-L DSC/OS197 TC analysis is performed for source terms consistent with heat load zone configuration (HLZC) 2. These source terms are much stronger than the HLZC 5 source terms developed for the 24PTH-S-LC DSC. The basket cross-section is also the same for the 24PTH-L and 24PTH-S-LC DSCs. Therefore, the 24PTH-L DSC/OS197 TC side dose rates presented in Tables P.5-3 and P.5-3a bound the side dose rates for the 24PTH-S-LC DSC within the OS197 TC.

The OS197 TC and Standardized TC have the same shielding dimensions and materials on the bottom and top (i.e., lid), although the OS197 TC is longer than the Standardized TC. Therefore, when the 24PTH-S-LC DSC is transferred in the OS197 TC, the Standardized TC bottom and top dose rates from Tables P.5-5 and P.5-5a remain bounding.

January 2023 Revision 2 72-1004 Amendment No. 18 Page P.5-24a P.5.4.8 Accident Models

No accident models were developed for the HSM-H because no accident scenario in Chapter P.11 has been identified that would alter the dose rates provided in Table P.5-1. For the HSM-Model 102 in an array, in an accident condition HSM-Model 102 is assumed to slide next to an adjacent HSM and therefore double the gap on one side as described in Chapter P.11. It is further conservatively assumed the dose rates from the array double as a result of this accident.

The HSM-Model 102 accident analysis and results are provided in Chapter P.11.

For both the OS197FC TC and Standardized TC, accident cases are performed assuming the neutron shield and steel neutron shield jacket (outer skin) of each have been torn off. The accident analysis for damaged fuels assumed as rubble is performed in Section U.5.4.8; the same conclusion is applicable for the 24PTH system. Therefore, the accident analysis performed with intact fuels is applicable to all fuel conditions. Accident dose rates at 1m, 100m, and 500m from the side of the cask are presented in Table P.5-3 and Table P.5-5 for the OS197FC TC and Standardized TC, respectively. Because accident dose rates are dominated by radiation exiting the side of the cask, accident dose rates computed for the Type 2 basket bound the Type 3 basket, as side dose rates decrease for the Type 3 basket.

P.5.4.9 OS197FC TC Models During Fuel Loading Operations

MCNP models are developed for the cask decontamination and welding operations during fuel loading using the 24PTH-L Type 2 and 3 DSCs. As the side and top dose rates from this cask with 24PTH-L DSC bounds the 24PTH-S-LC DSC due to the higher source term used in 24PTH-L DSC, calculations are not performed for the loading operations with Standardized TC with 24PTH-S-LC DSC.

Cask Decontamination. The 24PTH-L DSC and the OS197FC TC are assumed to be completely filled with water, including the region between 24PTH-DSC and cask, which is referred to as the cask/24PTH-DSC annulus. The 24PTH-DSC inner cover plate is assumed to be in place and the temporary shielding has not yet been installed. Results for this case are provided in Table P.5-4 and Table P.5-4a for the Type 2 and 3 baskets, respectively.

Welding and 24PTH-L DSC Draining. Before the start of welding operation, approximately 60% of the water in the DSC cavity is removed due to hydrogen generation. A dry DSC cavity is assumed in all welding models to be conservative. Temporary shielding consisting of three inches of NS3 and one inch of steel is assumed to cover the 24PTH-L DSC top shield plug. In addition, the DSC outer top cover plate is not present. The cask/24PTH-DSC annulus is assumed to remain completely filled with water. Results for this case are provided in Table P.5-4 and Table P.5-4a for the Type 2 and 3 baskets, respectively.

January 2023 Revision 2 72-1004 Amendment No. 18 Page P.5-24b Table P.5-14 Shielding Material Densities Assembly Region Material Densities Atomic Number Density (atom/b-cm)

Element Number Bottom End Fuel Plenum Top End Fitting Fitting O 8 - 1.35E-02 - -

Al 13 1.31E-05 3.61E-06 6.39E-05 2.98E-05 Ti 22 9.88E-06 2.72E-06 4.80E-05 2.24E-05 Cr 24 1.88E-03 6.62E-05 1.06E-03 2.99E-03 Mn 25 1.65E-04 - - 2.49E-04 Fe 26 5.96E-03 8.45E-05 1.29E-03 9.17E-03 Ni 28 1.21E-03 1.44E-04 2.54E-03 2.22E-03 Zr 40 6.23E-03 3.79E-03 3.89E-03 -

Mo 42 1.85E-05 5.08E-06 8.99E-05 4.19E-05 Sn 50 7.81E-05 4.75E-05 4.88E-05 -

U-235 92 - 3.39E-04 - -

U-238 92 - 6.37E-03 - -

Total 1.56E-02 2.43E-02 9.03E-03 1.47E-02

Other Shielding Materials

Number Density (atom/b-cm)

Element Atomic Number NS-3 Concrete Water Air Lead Carbon Steel Stainless Aluminum/B Steel ORAL

H 1 4.498E-02 7.767E-03 6.393E-02 B-10 5 3.054E-04 C 6 9.595E-03 N 7 3.587E-05 O 8 3.704E-02 4.317E-02 3.203E-02 9.534E-06 Na 11 1.022E-03 Al 13 6.887E-03 2.343E-03 6.071E-02 Si 14 1.243E-03 1.559E-02 K 19 6.776E-04 Ca 20 1.454E-03 2.855E-03 Cr 24 1.743E-02 Fe 26 1.042E-04 3.019E-04 8.465E-02 6.128E-02 Ni 28 7.511E-03 Pb 82 3.296E-02 Total 1.016E-01 7.373E-02 9.596E-02 4.540E-05 3.296E-02(1) 8.465E-02 8.622E-02 6.071E-02

Note:

(1) This correspond to a lead density of 11.34 g/cm3. Note that for the 24PTH-S-LC the Type 3 analysis, lead density employed in the shield plugs model is 11.00 g/cm3 (or 3.286E-2 atom/b-cm).

January 2023 Revision 2 72-1004 Amendment No. 18 Page P.5-78 P.9.1.5 Shielding Integrity Tests

The transfer cask poured lead shielding integrity will be confirmed via gamma scanning prior to first use. The detector and examination grid will be matched to provide coverage of the entire lead-shielded surface area. The acceptance criterion is attenuation greater than or equal to that of a test block matching the cask through-wall configuration with lead and steel thicknesses equal to the design minima less 5%.

The radial neutron shielding is provided by filling the neutron shield shell with water during operations. No testing is necessary. The neutron shield material in the lid and bottom end is a cementitious grout, NS-3. The shielding performance of this material will be assured by written procedures.

The gamma and neutron shielding materials of the storage system itself are limited to concrete HSM components and steel shield plugs in the DSC, except for the 24PTH-S-LC DSC. The integrity of these shielding materials is ensured by the control of their fabrication in accordance with the appropriate ASME, ASTM or ACI criteria. No additional acceptance testing is required.

The 24PTH-S-LC DSC incorporates lead in the top and bottom shield plugs, either installed by a pour or a precast insert. A volumetric inspection of the lead, either by a gamma inspection or an ultrasonic inspection, shall be performed to check for the existence and extent of any voids. The results of this inspection shall verify that the effective thickness of the lead through any section conforms to the thickness specified on the drawings.

P.9.1.6 Thermal Acceptance Tests

No thermal acceptance testing is required to verify the performance of each storage unit other than that specified in the Technical Specifications for initial loading.

The heat transfer analysis for the basket includes credit for the thermal conductivity of neutron-absorbing materials. Requirements for Type 1 and 2 baskets are specified in Section P.4.3. For Type 3 baskets, the minimum acceptable thermal conductivity is [

] Because these materials do not have publicly documented values for thermal conductivity, testing of such materials will be performed in accordance with Section P.9.1.7.6.

P.9.1.7 Poison Acceptance

The neutron absorber used for criticality control in the DSC basket may consist any of the following types of material:

a) Borated aluminum (Basket Types 1 and 2 only) b) Boron carbide/aluminum metal matrix composite (MMC) (All basket types) c) BORAL (Basket Types 1 and 2 only)

January 2023 Revision 2 72-1004 Amendment No. 18 Page P.9-2 The canister stop plates are loaded by the normal and off-normal handling loads and seismic loads. The normal handling load during the insertion of the DSC is 60 kips on both of the rails.

The maximum off-normal handling load is 80 kips on one rail. The seismic load considering a conservative factor of 1.5 is 95.625 kips acting on each plate. Stresses in the canister stop plates, rail-to-canister stop end plate weld, and canister stop end plate-to-stiffener plate welds are all determined to be less than the specified allowables.

R.3.7.8.8 Thermal Cycling of the HSM Model 152

No change to Section 8.2.10.5.

R.3.7.8.9 Evaluation of HSM Model 152 Concrete Components with Temperature Exceeding Code Limits

The maximum concrete temperature under normal and off-normal condition for the HSM Model 152 are 221ºF and 231/234ºF (for 117ºF and 125ºF ambient conditions), respectively. These temperatures exceed 200ºF in normal condition and 225ºF in off-normal condition, but do not exceed 300ºF. Therefore, as specified in CoC 1004 SER [3.9], no tests or reduction in concrete strength are required to demonstrate the capability of the concrete to adequately handle the elevated temperatures provided Type II cement is used and special aggregates are selected which are acceptable for concrete in this temperature range. In lie u of Type II cement, blended Portland cement meeting the requirements of ASTM C595 is acceptable for use. This a pproach is consistent with standardized HSM design, for which special aggregates for the roof concrete mix are provided.

The maximum concrete temperature for a 40-hour blocked vent condition is 394/397ºF (for 117ºF and 125ºF ambient conditions), which exceeds the 350ºF limit specified in CoC 1004 SER

[3.9]. As noted in the CoC 1004 SER [3.9], use of any Portland cement concrete where accident temperature exceeds 350ºF will require testing be performed on the exact concrete mix.

Portland cements meeting the requirements of ASTM C150 or ASTM C595 (blended Portland) are acceptable for use. Elevated temperature testing of the exact concrete mix (cement type, additives, water-cement ratio, aggregates, proportions) is to be performed for the HSM Model 152. The use of high temperature concrete testing is explicitly accepted by the NRC, as documented in the NRCs SER, Section 3.0, Page 3-5. The testing shall demonstrate the level of strength reduction is less than that which was applied (10% in the calculation), and show that the increased temperatures do not cause deterioration of the concrete.

January 2023 Revision 2 72-1004 Amendment No. 18 Page R.3-20

  • For off-normal and accident operating conditions, the maximum DSC shell temperatures for a 24P DSC stored in a standardized HSM Model 80/102 generally bound those computed for all DSC models stored in HSM Model 152. For the cases when the 32PT and 24PHB DSCs are stored within HSM Model 152, the maximum DSC shell temperatures are slightly higher (12 ºF) than those when the 32PT and 24PHB DSCs are in HSM Model 80/102.

The increase in DSC shell temperature is conservatively ~2% of absolute temperature. Thus based on gas laws the maximum increase in DSC internal pressure would be less than 2%. From the FSAR the bounding pressures for the DSC are for the Transfer Cask conditions (off-normal and accident), which are unchanged, and even in these cases there is more that 2% margin available.

R.4.4.2 Evaluation of HSM Model 152 Concrete Temperatures The maximum concrete temperatures for a 24 kW DSC stored in an HSM Model 152 are tabulated in Table R.4 -2 and are compared against the allowable concrete temperatures. It should be noted that the methodology used for calculating the HSM Model 152 concr ete temperature is the same as that used in [4.1] (HEATING7).

From a review of the results summarized in Table R.4 -2 below, the following observations can be made:

  • The maximum concrete temperatures for the normal and off -normal thermal conditions exceed 200 ºF but do not exceed 300 ºF, therefore no tests or reduction of concrete strength are required so long as Type II cement is used and special concrete aggregates are selected in accordance with the criteria recommended by the U.S. NRC in [4.3]. In lieu of Type II cement, blended Portland cement meeting the requirements of ASTM C595 is acceptable for use.
  • The maximum concrete temperature for the accident condition exceeds 350ºF, therefore testing is required on the exact concrete mix (cement type, additives, water -cement ratio, aggregates, proportions, etc.) to acceptably demonstrate the level of strength reduction which needs to be applied, and to show that the increased temperatures do not cause deterioration of the concrete. The use of high temperature concrete testing is acceptable, as documented in the SER [4.3], Section 3.0, Page 3-5. The testing shall demonstrate the level of strength reduction is less than that which was applied (10% in the calculation), and show that the increased temperatures do not cause deterioration of the concrete.

R.4.4.3 Evaluation of HSM Model 152 Maximum Fuel Cladding Temperatures A summary of the maximum fuel cladding temperatures evaluated for the various DSC models stored in HSM Model 152 is shown in Tab le R.4 -3. In all cases, calculated cladding temperatures are less than allowables.

R.4.4.4 Evaluation of HSM Model 152 Maximum Air Exit Temperatu re Table R.4 -4 documents the results of an evaluation that shows the equilibrium air temperature difference between the ambient temperature and the vent outlet temperature (T) with a 24 kW

January 2023 Revision 2 72-1004 Amendment No. 18 Page R.4-3 Table R.4-2 Maximum Concrete Temperatures for the HSM Model 152

Maximum 24kW DSC in HSM 152 HSM 152 Allowable Thermal Loading Ambient Concrete Temperature Concrete Temperature Condition Temperature (°F) (°F)

(°F)

Normal 100 221 300 (1)

Off-Normal 117 231 125 234 300 (1)

Accident 40 Hour 117 394 Blocked Vent 125 397 425 (2)

Notes:

1. Use of Type II cement in combination with special aggregates are selected which are acceptable for concrete in this temperature range as specified in the Discussion of Concrete Constituents and Temperature Suitability in the Safety Evaluation Report of Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel (Pages 3-4 and 3 -5), U.S. Nuclear Regulatory Commission, December 1994 [4.3]. In lieu of Type II cement, blended Portland cement meeting the requirements of ASTM C595 is acceptable for use.
2. Use of any Portland cement concrete where accident temperatures exceed 350 °F requires performance of tests on the exact concrete mix used as specified in the Safety Evaluation Report of Safety Anal ysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel (Page 3-5), U.S. Nuclear Regulatory Commission, December 1994 [4.3]. Portland cements meeting the requireme nts of ASTM C150 or ASTM C595 (blended Portland) are acceptable for use.

January 2023 Revision 2 72-1004 Amendment No. 18 Page R.4-6

20. Replace the transfer cask top cover plate. Secure the skid to the trailer, retract the vertical jacks and disconnect the skid positioning system.
21. If this is the final loading, fully drain the liquid neutron shield.
22. Tow the trailer and cask to the designated equipment storage area. Return the remaining transfer equipment to the storage area.
23. Close and lock the ISFSI access gate and activate the ISFSI security measures.
24. Not Used.

T.8.1.7 Monitoring Operations

1. Perform routine security surveillance in accordance with the licensees ISFSI security plan.
2. Perform one of the two alternate daily surveillance activities listed below:
a. A daily visual surveillance of the HSM air inlets and outlets to insure that no debris is obstructing the HSM ve nts in accordance wi th Technical Specification 4.3.6.a requirements.
b. A temperature measurement of the thermal performance, for each HSM, on a daily basis in accordance with Technical Specification 4.3.6.b requirements.

January 2023 Revision 2 72-1004 Amendment No. 18 Page T.8-13

14. Recheck all alignment marks and ready all systems for DSC transfer. The TC shall be aligned with respect to the HSM such that the longitudinal centerline of the DSC in the TC is within +/- inch of its true position when the TC is docked with the HSM front access opening.

If the alignment tolerance is exceeded, the following actions should be taken:

a. Confirm that the transfer system is properly configured,
b. Check and repair the alignment equipment, or
c. Confirm the locations of the alignment targets on the TC and HSM.
15. Activate the hydraulic ram to initiate insertion of the DSC into the HSM. Stop the ram when the DSC reaches the support rail stops at the back of the module.
16. Not used. 72.48
17. Retract and disengage the hydraulic ram system from the cask and move it clear of the cask.

Remove the cask restraints from the HSM.

18. Using the skid positioning system, disengage the cask from the HSM access opening.
19. Install the DSC axial in retainer through the HSM door opening.
20. Install the HSM door using a portable crane and secure it in place. Door may be welded for security. Verify that the HSM dose rates are compliant with the limits specified in CoC Appendix A Inspections, Tests, and Evaluations Items 3.3.1 and 3.3.2.
21. Replace the TC top cover plate. Secure the skid to the trailer, retract the vertical jacks and disconnect the skid positioning system.
22. If this is the final loading, fully drain the liquid neutron shield.
23. Tow the trailer and cask to the designated equipment storage area. Return the remaining transfer equipment to the storage area.
24. Close and lock the ISFSI access gate and activate the ISFSI security measures.
25. Not used.

U.8.1.7 Monitoring Operations

1. Perform routine security surveillance in accordance with the licensee's ISFSI security plan.
2. Perform one of the two alternate daily surveillance activities listed below:
a. A daily visual surveillance of the HSM air inlets and outlets to insure that no debris is obstructing the HSM vents in accordance with Technical Specification 4.3.6.a requirements.

January 2023 Revision 2 72-1004 Amendment No. 18 Page U.8-12 V.3.7.8.4 Evaluation of HSM Model 202 Support Steel The evaluation of the HSM Model 202 support steel is de scribed in Section P.3.7.11.6.4 since the HSM Model 202 is based on the HSM -H.

V.3.7.8.5 Evaluation of HSM Model 202 Shield Door The evaluation of the HSM Model 202 shield door is described in Section P.3.7.11.6.5 since the HSM Model 202 is based on the HSM -H.

V.3.7.8.6 Evaluation of HSM Model 202 Heat Shields The evaluation of HSM Model 202 heat shields is described in Section P.3.7.11.6.6 since the HSM Model 202 is based on the HSM -H.

V.3.7.8.7 Evaluation of HSM Model 202 Seismic Retainers The evaluation of HSM Model 202 seismic retainers is described in Section P.3.7.11.6.7 since the HSM Model 202 is based on the HSM -H.

V.3.7.8.8 Thermal Cycling of the HSM Model 202 No change to Section 8.2.10.5.

V.3.7.8.9 Evaluation of HSM Model 202 Concrete Components with Temperature Exceeding Code Limits The maximum concrete temperature under off-normal condition for the HSM Model 202 are 238/243 °F (for 117 °F and 125 °F ambient conditions). The normal condition is bounded by the off-normal condition. Although the maximum concrete temperatures exceed 225 °F in the off-normal condition, they do not exceed 300 °F. Therefore, as specified in [3.4], no tests or reduction in concrete strength are required to demonstrate the capability of the concrete to adequately handle the elevated temperatures provided Type II cement is used and special aggregates are selected which are acceptable for concrete in this temperature range. This approach is consistent with standardized HSM design, for which special aggregates for the roof concrete mix are provided. In lieu of Type II cement, blended Portland cement meeting the requirements of ASTM C595 is acceptable for use.

The maximum concrete temperature for a 40-hour blocked vent condition is 376/381 °F (for 117 °F and 125 °F ambient conditions), which exceeds the 350 °F limit specified in [3.4]. As noted in [3.4], use of any Portland cement concrete where accident temperature exceeds 350 °F will require testing be performed on the exact concrete mix. Elevated temperature testing of the exact concrete mix (cement type, additives, water-cement ratio, aggregates, proportions) is to be performed for the HSM Model 202. The use of high temperature concrete testing is explicitly accepted by the NRC, as documented in the NRCs SER [3.4], Section 3.0, Page 3-5. The testing shall demonstrate the level of strength reduction is le ss than that which was applied, and show that the increased temperatures do not cause deterioration of th e concrete. Portland cements meeting the requirements of ASTM C150 or ASTM C595 (blended Portland) are acceptable for use.

January 2023 Revision 2 72-1004 Amendment No. 18 Page V.3-12 Table V.4-2 Maximum Concrete Temperatures for the HSM Model 202

Maximum 24kW DSC in HSM 202 HSM 202 Allowable Thermal Loading Ambient Concrete Temperature Concrete Temperature Condition Temperature (°F) (°F)

(°F)

Normal 100 N/A(Bounded by Off-Normal) 300(1)

Off-Normal 117 238 125 243 300(1)

Accident 40 Hour 117 376 Blocked Vent 125 381 425(2)

Notes:

(1) Use of Type II cement in combination with special aggregates are selected which are acceptable for concrete in this temperature range as specified in the Discussion of Concrete Constituents and Temperature Suitability in the Safety Evaluation Report of Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel (Pages 3-4 and 3-5), U.S. Nuclear Regulatory Commission, December 1994 [4.3]. In lieu of Type II cement, blended Portland cement meeting the requirements of ASTM C595 are acceptable for use.

(2) Use of any Portland cement concrete where accident temperatures exceed 350 °F requires performance of tests on the exact concrete mix used as specified in the Safety Evaluation Report of Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel (Page 3-5), U.S.

Nuclear Regulatory Commission, December 1994 [4.3]. Portland cements meeting the requirements of ASTM C150 or ASTM C595 (blended Portland) are acceptable for use.

January 2023 Revision 2 72-1004 Amendment No. 18 Page V.4-6

21. Install the HSM door using a portable crane or other suitable lifting device and secure it in place. Door may be welded for security. Verify that the HSM dose rates are compliant with the limits specified in CoC Appendix A Inspections, Tests, and Evaluations Items 3.3.1 and 3.3.2.
22. Replace the TC top cover plate. Secure the skid to the trailer, retract the vertical jacks and disconnect the skid positioning system.
23. If this is the final loading, fully drain the liquid neutron shield.
24. Tow the trailer and cask to the designated equipment storage area. Return the remaining transfer equipment to the storage area.
25. Close and lock the ISFSI access gate and activate the ISFSI security measures.
26. Not Used.

W.8.1.7 Monitoring Operations

1. Perform routine security surveillance in accordance with the licensee's ISFSI security plan.
2. Perform one of the two alternate daily surveillance activities listed below:
a. A daily visual surveillance of the HSM air inlets and outlets to insure that no debris is obstructing the HSM vents in accordance with Technical Specification 4.3.6.a requirements.
b. A temperature measurement of the thermal performance, for each HSM, on a daily basis in accordance with Technical Specification 4.3.6.b requirements.

W.8.2 Procedures for Unloading the Cask

The operational differences specified above for loading operations when using OS197L TC (relative to the use of OS197 TC described in Chapter 5) will also apply for unloading operations.

W.8.3 Identification of Subjects for Safety Analysis

There is no change relative to Section 5.1.3 regarding criticality control, chemical safety, operational shutdown modes and maintenance techniques.

In addition to the typical instrumentation listed in Table 5.1-1 of Section 5.1.3, the use of OS197L TC shall require optical targets and instruments to implement specific remote crane operations described in Section W.8.1 above.

January 2023 Revision 2 72-1004 Amendment No. 18 Page W.8-25

22. Tow the trailer and cask to the designated equipment storage area. Return the remaining transfer equipment to the storage area.
23. Close and lock the ISFSI access gate and activate the ISFSI security measures.
24. Not used.

Y.8.1.7 Monitoring Operations

1. Perform routine security surveillance in accordance with the licensees ISFSI security plan.
2. Perform one of the two alternate daily surveillance activities listed below:
a. A daily visual surveillance of the HSM air inlets and outlets to ensure that no debris is obstructing the HSM vents in accordance with Technical Specification 4.3.6.a requirements.
b. A temperature measurement of the thermal performance, for each HSM, on a daily basis in accordance with Technical Specification 4.3.6.b requirements.

January 2023 Revision 2 72-1004 Amendment No. 18 Page Y.8-13

24. Close and lock the ISFSI access gate and activate the ISFSI security measures.
25. Not used.

Z.8.1.7 Monitoring Operations

1. Perform routine security surveillance in accordance with the licensee's ISFSI security plan.
2. Perform one of the two alternate daily surveillance activities listed below:
a. A daily visual surveillance of the HSM air inlets and outlets to ensure that no debris is obstructing the HSM vents in accordance with Technical Specification 4.3.6.a requirements.
b. A temperature measurement of the thermal performance, for each HSM, on a daily basis in accordance with Technical Specification 4.3.6.b requirements.

January 2023 Revision 2 72-1004 Amendment No. 18 Page Z.8-12a

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