ML19073A200
| ML19073A200 | |
| Person / Time | |
|---|---|
| Site: | 07201029 |
| Issue date: | 03/31/2019 |
| From: | Orano USA, TN Americas LLC |
| To: | Office of Nuclear Material Safety and Safeguards |
| Shared Package | |
| ML19073A204 | List: |
| References | |
| E-53756 ANUH-01.0150, Rev 9 | |
| Download: ML19073A200 (92) | |
Text
Enclosure 3 to E-53756 Replacement Pages for ANUH-01.0150, Standardized Advanced NUHOMS UFSAR, Revision 9 (Public Version).
NON-PROPRIETARY UPDATED FINAL SAFETY ANALYSIS REPORT FOR nrn STANDARDIZED ADVANCED NUHOMS HORIZONTAL MODULAR STORAGE SYSTEM FOR IRRADIATED NUCLEAR FUEL By TN Americas LLC(l)
Columbia, MD March 2019 ANUH-01.0150 Revision 9 (I) TN Americas LLC, formerly AREVA 1N, and Transnuclear, Inc. (herein referred to as 1N Americas LLC, AREVA 1N, Transnuclear, Inc., Transnuclear, or 1N)
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 REVISION LOG SHEET UFSAR Date Record of Changes/FCNs Changed Pages Revision 0
3/19/03 None All 1
3/21/05 FCNs 721029-39, 40, 62, 65, 81, 89, See List of Effective Pages 92, 124, 126, 165, 169 & 175 FCNs 721029-182, 185, 103 R-1, 162 2
8/17/06 R-1, 166, 173 R-1, 176 R-1,177 and See List of Effective Pages 204 3
8/15/08 FCNs 721029-202, 205,206,208,215, See List of Effective Pages 220,222 Rl, 232,239,246,257,272 4
8/12/10 FCNs 721029-275, 280 R-1, 285,294, See List of Effective Pages 303,311,312 R-1, 316 5
8/13/12 FCNs 721029-339, 348 R-1, 351 R-1, See List of Effective Pages 352,353,354,356,364 6
8/13/14 FCN 721029-385 See List of Effective Pages FCN 721029-374 R-1, 378 R-1, 7
8/11/16 386 R-1, 394,407 R-1, 414,415, See List of Effective Pages 416 R-1, 417 8
8/13/18 FCN 721029-418, 419 R-1, 420 R-1, See List of Effective Pages 421,422 9
3/12/19 FCN 721029-424 See List of Effective Pages ii
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 Revision 2 of this UFSAR incorporates changes implemented due to the approval of CoC 1029 Amendment 1, effective May 16, 2005. It also incorporates modifications implemented per 10 CFR 72.48 from March 21, 2005 through August 15, 2006.
Revision 3 of this UFSAR incorporates modifications implemented per 10 CFR 72.48 from August 16, 2006 through August 15, 2008. This revision also includes a full list of effective pages.
Revision 4 of this UFSAR incorporates modifications implemented per 10 CFR 72.48 from August 16, 2008 through August 12, 2010.
Revision 5 of this UFSAR incorporates modifications implemented per 10 CFR 72.48 from August 13, 2010 through August 13, 2012.
Revision 6 of this UFSAR incorporates modifications implemented per 10 CFR 72.48 from August 14, 2012 through August 13, 2014.
Revision 7 of this UFSAR incorporates changes implemented due to the approval of CoC 1029 Amendment 3, effective February 23, 2015. It also incorporates modifications implemented per 10 CFR 72.48 from August 14, 2014 through August 11, 2016.
Revision 8 of this UFSAR incorporates modifications implemented per 10 CFR 72.48 from August 12, 2016 through August 13, 2018.
Revision 9 of this UFSAR incorporates changes implemented due to the approval of CoC 1029 Amendment 4, effective March 12, 2019.
ANUH-01.0150 V
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I Pa!!e or description Rev.
Title Page 9
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LOEP-1 9
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08/18 DWG: (sh. 1 of 6) 61 7/18/18 NUH-05-4010 DWG: (sh. 2 of6) 61 Not shown NUH-05-4010 DWG: (sh. 3 of6) 61 Not shown NUH-05-4010 DWG: (sh. 4 of 6) 61 Not shown NUH-05-4010 DWG: (sh. 5 of 6) 61 Not shown NUH-05-4010 DWG: (sh. 6 of 6) 61 Not shown NUH-05-4010 DWG: (sh. 1 of9) 81 7/18/18 NUH-03-4011 DWG: (sh. 2 of9) 81 Not shown NUH-03-4011 DWG: (sh. 3 of9) 81 Not shown NUH-03-4011 DWG: (sh. 4 of9) 81 Not shown NUH-03-4011 DWG: (sh. 5 of9) 81 Not shown NUH-03-4011 1 Because SAR drawings were revised throughout the licensing period, their revision level may be higher than the overall UFSAR revision level.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9; 03/19 I List Of Effective Pages Page or description Rev.
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Advanced NUHOMS System Updated Final Safety Analysis Rep()r!_
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08/18 DWG: (sh. 1 of 8) i 7/18/18 ANUH-01-4001 DWG: (sh. 2 of 8) 72 Not shown ANUH-01-4001 DWG: (sh. 3 of 8) 73 Not shown ANUH-01-4001 1 Because SAR drawings were revised throughout the licensing period, their revision level may be higher than the overall UFSAR revision level.
2 Because SAR drawings were revised throughout the licensing period, their revision level may be higher than the overall UFSAR revision level.
ANUH-01.0150 LOEP-8 Page or description Rev.
Date DWG: (sh. 4 of 8) 74 Not shown ANUH-01-4001 DWG: (sh. 5 of 8) 7s Not shown ANUH-01-4001 DWG: (sh. 6 of 8) 76 Not shown ANUH-01-4001 DWG: (sh. 7 of 8) 77 Not shown ANUH-01-4001 DWG: (sh. 8 of 8) 78 Not shown ANUH-01-4001 A.2.1-1 5
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08/06 3 Because SAR drawings were revised throughout the licensing period, their revision level may be higher than the overall UFSAR revision level.
4 Because SAR drawings were revised throughout the licensing period, their revision level may bt, higher than the overall UFSAR revision level.
5 Because SAR drawings were revised throughout the licensing period, their revision level may be higher than the overall UFSAR revision level.
6 Because SAR drawings were revised throughout the licensing period, their revision level may be higher than the overall UFSAR revision level.
7 Because SAR drawings were revised throughout the licensing period, their revision level may be higher than the overall UFSAR revision level.
8 Because SAR drawings were revised throughout the licensing period, their revision level may be higher than the overall UFSAR revision level.
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08/16 ANUH-01.0150 LOEP-19
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08/16 ANUH-01.0150 LOEP-20
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08/16 ANUH-01.0150 LOEP-21
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08/16 ANUH-01.0150 LOEP-22
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08/16 ANUH-01.0150 LOEP-23
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I 3.5.3.3 WE 14X14 (MOX) Fuel Using the geometric and material properties in Table 3.5-4 through Table 3.5-6 and the methodology in Section 3.5.3.1, analysis of the MOX fuel Zircaloy-4 clad fuel assemblies for 75g side and 25g comer drops gives the following results. The side drop allowable g-loading is calculated to be 165g which exceeds the postulated 75g load. For the comer drop, the critical buckling load is calculated to be 71.5g which, when combined with the side drop component, results in a ratio of 0.32. This provides a factor of safety of greater than 3 against fuel rod failure in a comer drop.
3.5.3.4 Results The fuel cladding for both the WE 14x14 stainless steel clad and Zircaloy-4 clad MOX assemblies will maintain structural integrity for both side and comer drop events.
3.5.4 Fuel Unloading For unloading operations during the time period when the spent fuel pool is available, the 24PT1-DSC will be filled with spent fuel pool water through the siphon port. During this filling operation, the 24PT1-DSC vent port is maintained open with effluents routed to the plant's off-gas monitoring system. The NUHOMS operating procedures recommend that the 24PT1-DSC cavity atmosphere be sampled first before introducing any reflood water in the 24PT1-DSC cavity.
When the pool water is added to a 24PT1-DSC cavity containing hot fuel and basket components, some of the water will flash to steam causing internal cavity pressure to rise. This steam pressure is released through the vent port. The procedures also specify that the flow rate and temperatures of the reflood water be controlled to ensure that the internal pressure in the 24PT1-DSC cavity is maintained at less than or equal to 20 psig. The reflood for the 24PT1-DSC is considered as a Service Level D event. The 24PT1-DSC is also evaluated for a Service Level D pressure of 60 psig. Therefore, there is sufficient margin in the 24PT1-DSC internal pressure during the reflooding event to assure that the 24PT1-DSC will not be over pressurized.
The maximum fuel cladding temperature during the reflooding will be significantly less than the vacuum drying condition due to the presence of water/steam in the 24PT1-DSC cavity. The analysis presented in Chapter 4 shows that the maximum cladding temperature during steady state vacuum drying operation is 751 °F. Therefore, the maximum cladding temperature during the reflooding operation will be less than 751 °F. This is still considerably below the short term cladding temperature limit of 806 °F. Therefore, no cladding damage is expected due to the reflood event. This is also substantiated by the operating experience gained with the loading and unloading of transportation packages like the IF-300 [3.35] which show that fuel cladding integrity is maintained during these operations and fuel handling and retrieval is not impacted.
ANUH-01.0150 3.5-5 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I Table 4.4-12 Technical Specifications 5.2.5.b Temperature Monitoring Limits for the 24PT1-DSC Max Temp (
0F)
Max Temp Rise (°F)
(in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)
Single Thermocouple (y = 34.5", x = 0, z = 4.75")
225 801 Dual Thermocouple (y = 60", x = +/-15", z = -11.25")
175 82
- 1.
Based on a 24 kW DSC heat load, as noted in Technical Specification Section 5.2.5.b at the analyzed location in the AHSM base.
- 2.
Based on a 14 kW DSC heat load, at the dual "as-built" thermocouple locations provided in the AHSM roof. A limit of 3 °F applies if the surveillance period is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> instead of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ANUH-01.0150 4.4-25a All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9. 03/19 I 4.7 Thermal Evaluation for Loading/Unloading Conditions All individual fuel assembly transfer operations occur when the 24PT1-DSC is in the spent fuel pool. The fuel is always submerged in free-flowing pool water permitting heat dissipation.
After fuel loading is complete, the 24PT1-DSC is removed from the pool, drained, dried, and backfilled with helium.
The two loading conditions evaluated for the Advanced NUHOMS System are the heatup of the 24PT1-DSC before its cavity can be backfilled with helium and the vacuum drying transient.
Transient thermal analyses are performed to predict the heatup time history for the 24PT1-DSC components during these events.
The unloading operation considered is the reflood of the 24PT1-DSC with water.
4.7.1 Vacuum Drying Thermal Analysis Analyses were performed for the vacuum drying condition in order to ensure that the fuel cladding and 24PT1-DSC structural component temperatures remain below the maximum allowable limits shown in Table 4.7-1. For every component except the spacer disc, steady state temperature distributions gave satisfactory results. To show compliance with the ASME B&PV Code [ 4. 7] temperature limits for the spacer disc material, transient analyses were performed to determine the time to reach 700°F, the temperature limit for SA-537, Class 2 plate. These time limits for the vacuum drying case are shown in Table 4.7-2.
For the steady state analysis, the model is similar to the model described in Section 4.4.2.5 and shown in Figure 4.4-6, Figure 4.4-7, and Figure 4.4-8. The exception is that the helium regions are replaced with air. Assuming that the cavity is filled with air during the vacuum drying operation provides conservative results since during the majority of the vacuum drying operation, the 24PT1-DSC cavity void volume is filled with a mixture of air, water and water vapor, and no credit is taken for evaporation of water, which is a strong cooling mechanism that takes place during this operation. Air thermal conductivity does not change significantly at lower pressures, therefore, the use of a thermal conductivity for a pressure higher than 3 Torr is acceptable. In accordance with Chapter 8, water is required to be in the annulus between the 24PT1-DSC and the transfer cask during the vacuum drying process. Therefore, the 24PT1-DSC shell boundary is set to a temperature of 230°F as a conservative estimate of the shell wall temperature during this operation. A heat load of 14 kW is considered in computing the maximum fuel cladding temperature. The 14 kW heat load is also used to calculate the maximum 24PT1-DSC component temperatures. The resulting maximum temperatures are tabulated in Table 4.4-6 and Table 4.4-7 for the basket structural components and fuel cladding respectively.
For the transient analysis, the model from Section 4.4.4 is used with the constant temperature boundary condition described above and the change to the helium regions described above. The density and specific heat of the basket materials and fuel assembly from Section 4.2 are also used in the HEATING? model. The time transient is measured from the beginning of the blowdown procedure to the beginning of the final helium backfill procedure. Therefore, the initial temperature of the basket is conservatively set to the saturation temperature of water as an initial condition. The transient vacuum drying case is performed for heat loads of 13 and 14 kW.
ANUH-01.0150 4.7-1 Al_l changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 The results of the transient analysis are presented in Table 4.7-2 and Figure 4.7-1. The resulting time limitations are incorporated into Chapter 12.
4.7.2 Pressure During Unloading of Cask To unload the fuel from the 24PT1-DSC during the time period when the spent fuel pool is available, reflooding of the 24PT1-DSC cavity is required. This occurs by first reducing the pressure in the 24PT1-DSC to atmospheric conditions followed by introducing water into the 24PT1-DSC through the drain port and venting through the vent port. Since fuel temperatures are expected to be significantly higher than the saturation temperature of water, flooding of the hot 24PT1-DSC will result in steam being generated which, if not vented, will result in a higher cavity pressure.
The flow rate of water into the 24PT1-DSC during reflood is controlled during this operation such that the pressure within the 24PT1-DSC stays below the design pressure of 20 psig for this condition.
4.7.3 Cask Heatup Analysis Heatup of the water within the 24PT1-DSC cavity prior to blowdown and backfilling with helium occurs as operations are being performed to decon the cask and drain and dry the 24PT1-DSC. Prevention of boiling in the Advanced NUHOMS System is not required to ensure public health and safety for the following reasons:
- 1.
The criticality analysis already considers a wide range of moderator densities which include that of steam (Chapter 6). Criticality limits were shown to be met even at conditions of low moderator density (boiling water).
- 2.
The cavity is always vented during the water heatup transient.
- 3.
Although steam may be produced through boiling of the water in the 24PT1-DSC, its presence in the weld joint area during inner cover plate installation operations will be
- essentially blocked at the interface between the shield plug and the support ring. What little steam that may be present is displaced by the argon shielding gas used in the GTAW process. This shielding gas is heavier than air (and steam) and is delivered at a sufficiently high rate (usually 30 - 50 ft3/hr) to assure that the steam will be effectively excluded from the weld joint. Finally, if moisture somehow did enter the weld area, the resulting weld bead porosity would be readily detectable by the visual inspection of each pass performed by the welding operator and the dye penetrant (PT) examination performed on the surface of the root pass.
Therefore, the only potential concern associated with steam generation is shielding. An unexpectedly high loss of water within the 24PT1-DSC cavity during these loading operations could result in increased occupational exposure. The following analysis is presented to identify to the license holders the time for the water in the 24PT1-DSC cavity to boil so that corrective action can be planned and implemented as necessary to address ALARA concerns.
ANUH-01.0150 4.7-2 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Surface Rear<1>
Fronf3J Roof Side11>
Table 5.1-2 Summary AHSM Dose Rates Dose Rate Maximum Dose Minimum Dose Component Rate, mrem/hr Rate, mrem/hr Gamma 0.11 1.06E-04 Neutron 0.01 3.04E-06 Gamma 45.27 1.59E-02 Neutron 0.54 6.99E-04 Gamma 3.57 3.57E-04 Neutron 0.05 2.37E-05 Gamma 1.35 5.13E-06 Neutron 0.03 4.61E-08 (1)
Rear and side does rates are on the outer surfaces of the shield walls.
Rev. 9, 03/19 I Average Surface Dose Rate<2>,
mrem/hr 4.06E-03 3.?0E-04 1.89 0.04 0.03 8.56E-04 0.26 0.01 (2)
These dose rates are bounding for 1 meter occupational exposures during transfer operations.
(3)
The maximum dose rates on the front surface are based on the results calculated in front of the entrance of the bottom air inlet. Around the door centerline (between 152 cm to 394 cm above the ground) has maximum dose rate 4.04 mremlhr (gamma dose rate 3.97 mremlhr and neutron dose rate 0.07 mremlhr).
ANUH-01.0150 5.1-3 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 8.2.2 Removal of Fuel from the 24PT1-DSC When the 24PT1-DSC has been removed from the AHSM, there are several potential options for off-site shipment of the fuel. These options include, but are not limited to, shipping the 24PT1-DSC with fuel assemblies or removing the fuel from the 24PT1-DSC as described below.
It is preferred to ship the 24PT1-DSC intact to a reprocessing facility, monitored retrievable storage facility or permanent geologic repository in a compatible shipping cask, such as the MP187, licensed under 10 CFR Part 71. However, there are several reasons why it may be necessary to remove fuel assemblies from the 24PT1-DSC during the time period when the spent fuel pool is available. These include off-site transport in a transport cask requiring an alternate canister configuration, return of fuel assemblies to a spent fuel pool, or placement of fuel assemblies in a different 24PT1-DSC. Other reasons might include removing fuel assemblies at the end of service life or for inspection following an accident as discussed in Chapter 12.
If it becomes necessary to remove fuel from the 24PT1-DSC prior to off-site shipment, there are two basic options available at the ISFSI or reactor site. The fuel assemblies could be removed and reloaded into a shipping cask using dry transfer techniques, or if the applicant so desires, the initial fuel loading sequence could be reversed and the plant's spent fuel pool utilized, if available. Procedures for unloading the 24PT1-DSC in a fuel pool are presented here, however wet or dry unloading procedures are essentially identical to those of24PT1-DSC loading through the weld removal process (beginning of preparation to placement of the transfer cask in the fuel pool). Prior to opening the 24PT1-DSC, the following operations are to be performed.
- 1.
The transfer cask may now be transferred to the cask handling area inside the plant's fuel handling building.
- 2.
Position and ready the trailer for access by the crane.
- 3.
Attach the lifting yoke to the crane hook.
- 4.
Engage the lifting yoke with the trunnions of the transfer cask.
- 5.
Visually inspect the yoke lifting hooks to insure that they are properly aligned and engaged onto the transfer cask trunnions.
- 6.
Lift the transfer cask approximately one inch off the trunnion supports. Visually inspect the yoke lifting hooks to insure that they are properly positioned on the trunnions.
- 7.
Move the crane in a horizontal motion while simultaneously raising the crane hook vertically and lift the transfer cask off the trailer. Move the transfer cask to the cask decontamination area.
- 8.
Lower the transfer cask into the cask decontamination area in the vertical position.
ANUH-01.0150 8.2-2 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19
- 9.
Wash the transfer cask to remove any dirt which may have accumulated during the 24PT1-DSC unloading and transfer operations.
- 10.
Place scaffolding around the transfer cask so that any point on the surface of the transfer cask is easily accessible to handling personnel.
- 11.
Unbolt the transfer cask top cover plate.
- 12.
Connect the rigging cables to the transfer cask top cover plate and lift the cover plate from the transfer cask. Set the transfer cask cover plate aside and disconnect the lid lifting cables.
- 13.
Install temporary shielding to reduce personnel exposure as required. Fill the transfer cask/24PT1-DSC annulus with clean demineralized water and seal the annulus.
The process of unloading the 24PT1-DSC into the spent fuel pool, if available, is similar to that used for loading. Operations that involve opening the 24PT1-DSC described below are to be carefully controlled in accordance with plant procedures. These operations are to be performed under the site's standard health physics guidelines for welding, grinding, and handling of potentially highly contaminated equipment. These are to include the use of prudent housekeeping measures and monitoring of airborne particulates. Procedures may require personnel to perform the work using respirators or supplied air.
If fuel needs to be removed from the 24PT1-DSC, precautions must be taken for the presence of damaged or oxidized fuel and to prevent radiological exposure to personnel during this operation. If degraded fuel is suspected, additional measures appropriate for the specific conditions are to be planned, reviewed, and implemented to minimize exposures to workers and radiological releases to the environment. A sampling of the atmosphere within the 24PT1-DSC should be taken prior to inspection or removal of fuel.
If the work is performed outside the fuel handling building, a tent may be constructed over the work area which may be kept under a negative pressure to control airborne particulates. Any radioactive gas release will be Kr-85, which is not readily captured. Whether the krypton is vented through the plant stack or allowed to be released directly depends on the plant operating requirements.
Following opening of the 24PT1-DSC, it is filled with demineralized or pool water prior to placement in the spent fuel pool to prevent a sudden inrush of pool water. Parameters related to reflooding the 24PT1-DSC cavity are addressed in Chapter 3. Place transfer cask into the pool.
The fuel unloading procedures listed below will be governed by the plant operating license under 10 CFR Part 50, and assume the availability of the spent fuel pool. The generic procedures for these operations are as follows:
- 1.
Locate the siphon and vent port using the indications on the top cover plate. Place a portable drill press on the top of the 24PT1-DSC. Align the drill over the siphon port.
ANUH-01.0150 8.2-3 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 exposure of personnel in the vicinity. The actual local and off-site dose rates, recovery time and operations needed to retrieve the cask, and the required actions to be performed following the event, depend upon the severity of the event, site characteristics and the resultant cask and trailer/skid damage.
11.2.5.4 Corrective Actions The DSC and transfer cask will be inspected for damage.
For recovery of the cask and contents, it may be necessary to develop a special sling/lifting apparatus to move the transfer cask. This may require several weeks of planning to ensure all steps are correctly organized. During this time, lead blankets may be added to the transfer cask to minimize on-site exposure to site operations personnel. The transfer cask would be roped off to ensure the safety of the site personnel.
The recovery operations listed in this section assume the cask drop occurs during initial transfer and loading of the DSC into the AHSM, when the spent fuel pool is still operational and available. If a drop of the transfer cask with a loaded DSC occurs during transfer to a transportation cask and an inspection determines that the DSC is damaged and a spent fuel pool is not available onsite, the DSC shall be placed into a safe condition. If required, the DSC could be transported offsite to a site licensed for either dry or wet unloading of the DSC.
11.2.6 Lightning 11.2.6.1 Cause of Accident Lightning striking the AHSM and causing an off-normal condition is not considered credible.
Lightning protection system requirements are site specific and depend upon the frequency of occurrence of lightning storms* in the proposed ISFSI location and the degree of protection offered by other grounded structures in the proximity of the AHSMs. The addition of simple lightning protection equipment, if required by plant criteria, to AHSM structures (i.e., grounded handrails, ladders, etc.) is considered a miscellaneous attachment and is allowed by the AHSM drawing (Dwg. No. NUH-03-4011), Section 1.5.2.
11.2.6.2. Accident Analysis Should lightning strike in the vicinity of the AHSM the normal storage operations of the AHSM will not be affected. The current discharged by the lightning will follow the low impedance path offered by the surrounding structures. Therefore, the AHSM will not be damaged by the heat or mechanical forces generated by current passing through the higher impedance concrete. Since the AHSM requires no electrical equipment for its continued operation, the resulting current surge from the lightning will not affect the normal operation of the AHSM.
11.2.6.3 Accident Dose Calculations Since no off-normal condition will develop as the result of lightning striking in the vicinity of the AHSM, no radiological consequences are expected.
ANUH-01.0150 11.2-27 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 12.3.0 Limiting Condition for Operation (LCO) and Surveillance Requirements (SR)
Applicability BASES LCOs LCO 3.0.1 LCO 3.0.2 ANUH-01.0150 LCO 3.0.1, 3.0.2, 3.0.4 and 3.0.5 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.
LCO 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LCO is required to be met (i.e.,
when the canister is in the specified conditions of the Applicability statement of each Specification).
LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered. The Required A~tions establish those remedial measures that must be taken within specified Completion Times when the requirements of an LCO are not met. This Specification establishes that:
- a.
Completion of the Required Actions within the specified Completion Times constitutes compliance with a Specification; and
- b.
Completion of the Required Actions is not required when an LCO is met within the specified Completion Time, unless otherwise specified.
There are two basic types of Required Actions. The first type of Required Action specifies a time limit in which the LCO must be met. This time limit is the Completion Time to restore a system or component or to restore variables to within specified limits. If this type of Required Action is not completed within the specified Completion Time, the canister may have to be placed in the spent fuel pool, if available, and unloaded. (Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS.) The second type of Required Action specifies the remedial measures that permit continued operation of the unit that is not further restricted by the Completion Time. In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation.
Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications.
12-5 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I
- 14. DECOMMISSIONING This chapter addresses decommissioning for the NUHOMS 24PT1 system. Decommissioning for the NUHOMS 24PT4 system is addressed in Appendix A, Chapter A.14. Decommissioning for the NUHOMS 32PTH2 system is addressed in Appendix B, Chapter B.14.
14.1 Decommissioning Considerations The Advanced NUHOMS System design features inherent ease and simplicity for decommissioning by providing easily decontaminable surfaces and isolating the external surfaces of the 24PT1-DSC from contact with the fuel pool. At the end of its service life, the 24PT1-DSC decommissioning could be performed by one of the options listed below:
Option 1, the 24PT1-DSC, including stored spent fuel, could be shipped to either a monitored retrievable storage system (MRS) or a geological repository for final disposal, or Option 2, the spent fuel could be removed from the 24PT1-DSC in the spent fuel pool, if still available onsite, or using dry transfer techniques or other means, and the fuel shipped ojfsite in an NRC approved transportation cask.
The first option requires that the 24PT1-DSC be upgraded to current Part 71 regulations. An amendment to C of C 71-9255 [14.2] has been approved by the NRC to allow for transport.of this 24PT1-DSC using the MP187 cask.
The first option does not require any decommissioning of the 24PT1-DSC. No residual contamination is expected to be left behind on the concrete AHSM. The AHSM, fence, and peripheral utility structures will require no decontamination or special handling after the last 24PT1-DSC is removed. The AHSM, fence, and peripheral utility structures could be -
demolished and recycled with normal construction techniques.
The second option, which assumes the availability of a spent fuel pool onsite, would require decontamination of the 24PT1-DSC and transfer cask (if applicable). The sources of contamination in the interior of the 24PT1-DSC or transfer cask would be the primary contamination left from the spent fuel pool water if unloading using the spent fuel pool; or crud, hot particles and fines from the spent fuel pins. This contamination could be removed with a high pressure water spray. If further surface decontamination of the 24PT1-DSC or transfer cask is necessary, electropolishing or chemical etching can be used to clean the contaminated surface.
After decontamination, the 24PT1-DSC and/or transfer cask could be cut up for scrap, partially scrapped, or refurbished for reuse. Any activated metal would be shipped as low level.
radioactive waste to a near surface disposal facility.
A review of cask activation analyses previously performed for similar systems (TN-32 cask
[14.4] and NUHOMS site license storage system) indicates that the levels of activation of the 24PT1-DSC, AHSM and transfer cask would be orders of magnitude below the specific activity of the isotopes listed in Tables 1 and 2 of 10 CFR 61.55 [14.3]. A detailed analysis is not considered necessary based on the significant margins determined from these analyses. A comparison of the source terms for this application to those referenced above including the activation analysis summary for the above applications is provided below:
ANUH-01.0150 14.1-1 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I Comparison of Source Terms for Activation Analyses Source Term NUHOMS Site (including Control 24PT1-DSC TN-32 (Metal Cask)
Comoonents}
License HSM y (y/sec/assy) 3.4 X 1015 5.3 x.1015 1.53 X 1015 n (n/sec/assy) 2.8 X 108 3.3 X 108 2.23 X 108 TN 32 and NUHOMS Site License HSM Activation Analysis Results Activity Ci/m 3
Nuclide HSM 10 CFR 61.55 Concrete HSM Steel TN-32 Limit H-3 8.3 X 10-11 40 C-14 2.3 X 10-iu 8
Co-60 4.4 X 10-b 8.1 X 10-£ 7.7 X 10-0 700 Ni-59 1.4 X 10-,u 3.1 X 10-6 2.5x10:o 220 Ni-63 8.3 X 10-1:l 3.2x10-4 3.4 X 10 4
3.5 Nb-94 3.9 X 10-!:l
.2
<5 year 4.6 X 10-3 2.0x10-1 2.3 X 10-2 700 half life Following surface decontamination, the radiation levels in the 24PT1-DSC or transfer cask due to activation will be below the acceptable limits of Regulatory Guide 1.86 [14.1]. The activation levels of the 24PT1-DSC or transfer cask materials will be far below the specific activity limits for both short and long lived nuclides for Class A waste. A detailed evaluation will be performed at the time of decommissioning to determine the appropriate mode of disposal, should refurbishment not be elected.
The procedure for decommissioning a 24PT1-DSC or transfer cask not being returned to service is summarized below:
Remove fuel in accordance with the unloading procedures of Chapter 8.
Survey interior of 24PT1-DSC or transfer cask. If the spent fuel pool is available, wash down the inside of the 24PT1-DSC or transfer cask. Pump out and filter contaminated water and cleaning agent. Survey interior of 24PT1-DSC or transfer cask again, decontaminate as required. It is expected that surface decontamination will be minimal.
If so, dispose of the 24PT1-DSC or transfer cask body as scrap metal. If unable to decontaminate to acceptable levels, the 24PT1-DSC and/or transfer cask body can be disposed of as low level radioactive waste.
Decontaminate the top inner and outer cover plates until able to dispose of as scrap metal.
If unable to achieve acceptable levels, dispose of them as low level radioactive waste.
The fuel unloading and decontamination steps for 24PT1-DSC, AHSM, or cask refurbishment are as outlined for the scrap choices, discussed above. However, the only pieces discarded are components damaged by unloading or that are considered to be difficult to decontaminate.
Following a comprehensive survey to confirm continued 24PT1-DSC, AHSM or transfer cask ANUH-01.0150 14.1-2 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9. 03/19 j A.I A.2 TABLE OF CONTENTS Page GENERAL INFORMATION......................................... :........................................................ A.1.1-1 A.1.1 Introduction............................................................................................................. A.1.1-2 A.1.2 General Description of the Advanced NUHOMS System.................................... A.1.2-1 A.1.3 A.1.4 A.1.5 A.1.2.1 Advanced NUHOMS System Chara.cteristics.................................. A.1.2-1 A.1.2.2 Operational Features.......................................................................... A.1.2-2 A.1.2.3 24PT4-DSC Contents......................................................................... A.1.2-3 Identification of Agents and Contractors................................................................ A.1.3-1 Generic Cask Arrays............................................................................................... A.1.4-1 Supplemental Data.................................................................................................. A.1.5-1 A.1.5.1 References.......................................................................................... A.1.5-1 A.1.5.2 Drawings............................................................................................ A.1.5-1 PRINCIPAL DESIGN CRITERIA......................................................................................... A.2.1-1 A.2.1 Spent Fuel to be Stored........................................................................................... A.2.1-1 A.2.2 A.2.3 A.2.4 A.2.5 A.2.6 A.2.1.1 Detailed Payload Description............................................................. A.2.1-1 Design Criteria for Environmental Conditions and Natural Phenomena................ A.2.2-1 A.2.2.1 Tornado and Wind Loadings.............................................................. A.2.2-1 A.2.2.2 Water Level (Flood) Design.............................................................. A.2.2-1 A.2.2.3 Seismic Design................................................................................... A.2.2-2 A.2.2.4 Snow and Ice Loadings...................................................................... A.2.2-2 A.2.2.5 Tsunami.............................................................................................. A.2.2-2 A.2.2.6 Lightning............................................................................................ A.2.2-2 A.2.2. 7 Combined Load Criteria..................................................................... A.2.2-2 A.2.2.8 Burial Urtder Debris........................................................................... A.2.2-3 A.2.2.9 Thermal Conditions............................................................................ A.2.2-3 Safety Protection Systems....................................................................................... A.2.3-1 A.2.3.1 General............................................................................................... A.2.3-1 A.2.3.2 Protection by Multiple Confinement Barriers and Systems............... A.2.3-1 A.2.3.3 Protection by Equipment and Instrumentation Selection................... A.2.3-3 A.2.3.4 Nuclear Criticality Safety................................................................... A.2.3-3 A.2.3.5 Radiological Protection.............................. ;....................................... A.2.3-4 A.2.3.6 Fire and Explosion Protection............................................................ A.2.3-5 A.2.3.7 Acceptance Tests and Maintenance................................................... A.2.3-5 Decommissioning Considerations........................................................................... A.2.4-1 Structures, Systems and Components Important to Safety..................................... A.2.5-1 A.2.5.1 Dry Shielded Canister
............................................. A.2.5-1 A.2.5.2 Advanced Horizontal Storage Module............................................... A.2.5-1 A.2.5.3 ISFSI Basemat and Approach Slabs.................................................. A.2.5-1 A.2.5.4 Transfer Equipment............................................................................ A.2.5-2 A.2.5.5 Auxiliary Equipment.......................................................................... A.2.5-2 Supplemental Information....................................................................................... A.2.6-1 A.2.6.1 References.......................................................................................... A.2.6-1 ANUH-01.0150
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9. 03/19 I A.3 A.4 STRUCTURAL EVALUATION............................................................................................ A.3.1-1 A.3.1 Structural Design..................................................................................................... A.3.1-1 A.3.2 A.3.3 A.3.4 A.3.5 A.3.6 A.3.1.1 Discussion.......................................................................................... A.3.1-1 A.3.1.2 24PT4-DSC and AHSM Design Criteria.,......................................... A.3.1-3 Weights and Centers of Gravity.............................................................................. A.3.2-1 Mechanical Properties of Materials......................................................................... A.3.3-1 A.3.3.1 24PT4-DSC Material Properties........................................................ A.3.3-1 A.3.3.2 AHSM Material Properties................................................................ A.3.3-2 A.3.3.3 Materials Durability........................................................................... A.3.3-2 General Standards for 24PT4-DSC and AHSM...................................................... A.3.4-1 A.3.4.1 Chemical and Galvanic Reactions..................................................... A.3.4-1 A.3.4.2 Positive Closure................................................................................. A.3.4-1 A.3.4.3 Lifting Devices................................................................................... A.3.4-2 A.3.4.4 Heat.................................................................................................... A.3.4-2 A.3.4.5 Cold.................................................................................................... A.3.4-4 Fuel Rods General Standards for 24PT4-DSC........................................................ A.3.5-1 A.3.5.1 Fuel Rod Temperature Limits for Westinghouse-CENP, Combustion Engineering 16x16 Fuel........................... ~..................... A.3.5-1 A.3.5.2 Fuel Assembly Thermal and Irradiation Growth............................... A.3.5-1 A.3.5.3 Fuel Rod Integrity During Drop Scenario.......................................... A.3.5-2 A.3.5.4 Fuel Unloading................................................................................... A.3.5-2 Supplemental Data.................................................................................................. A.3.6-1 A.3.6.1 24PT4-DSC Structural Analysis........................................................ A.3.6-1 A.3.6.2 Structural Analysis of the AHSM.................................................... A.3.6-17 A.3. 7 References............................................................................................................... A.3. 7-1 THERMAL EVALUATION...................................................................................................... A.4-1 A.4.1 Discussion............ :.................................................................................................. A.4.1-1 A.4.2 A.4.3 A.4.4 A.4.5 A.4.6 A.4.1.1 Overview and Purpose of Thermal Analysis..................................... A.4.1-1 A.4.1.2 Thermal Load Specification/Ambient Temperature.......................... A.4.1-1 Summary of Thermal Properties ofMaterials......................................................... A.4.2-1 Specifications for Components.............,................................................................. A.4.3-1 Thermal Evaluation for Normal and Off-Normal Conditions of Storage and Transfer................................................................................................................... A.4.4-1 A.4.4.1 Overview of Thermal Analysis for Normal and Off-Normal A.4.4.2 A.4.4.3 A.4.4.4 A.4.4.5 A.4.4.6 A.4.4.7 A.4.4.8 A.4.4.9 A.4.4.10 Conditions of Storage and Transfer................................................... A.4.4-1 Thermal Analysis of24PT4-DSC in the AHSM................................ A.4.4,-1 Thermal Analysis of 24PT4-DSC in the TC...................................... A.4.4-5 24PT4-DSC Basket Thermal Analysis............................................... A.4.4-6 Test Model....................................................................................... A.4.4-10 Maximum Temperatures.................................................................. A.4.4-10 Minimum Temperatures................................................................... A.4.4-10 Maximum Internal Pressure............................................................. A.4.4-10 Maximum Thermal Stresses............................................................. A.4.4-11 Evaluation of System Performance for Normal Conditions of Storage and Transfer........................................................................ A.4.4-12 Thermal Evaluation for Off-Normal Conditions..................................................... A.4.5-1 Thermal Evaluation for Accident Conditions......................................................... A.4.6-1 ANUH-01.0150 ii
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9. 03/19 I A.4.7 A.4.8 A.4.9 A.4.10 A.4.6.1 A.4.6.2 A.4.6.3 A.4.6.4 A.4.6.5 A.4.6.6 A.4.6.7 Accident Ambient Conditions............................................................ A.4.6-1 Blockage of AHSM Inlet and Outlet Vents....................................... A.4.6-1 TC Loss of Neutron Shield and Sunshade......................................... A.4.6-4 Fire Accident Evaluation................................................................... A.4.6-4 Flood Accident................................................................................... A.4.6-5 Maximum Pressure............................................................................ A.4.6-5 Evaluation of System Performance for Accident Conditions of Storage and Transfer.......................................................................... A.4.6-5 Thermal Evaluation for Loading/Unloading Conditions......................................... A.4.7-1 A.4.7.1 Vacuum Drying Thermal Analysis.................................................... A.4.7-1 A.4.7.2 Pressure during Unloading of Cask.................................................... A.4.7-5 A.4.7.3 Cask Heatup Analysis........................................................................ A.4.7-5 A.4.7.4 Pressure During Loading of Cask...................................................... A.4.7-6 Confirmatory Thermal Analysis of the 24PT4-DSC............................................... A.4.8-l Determination of Effective Thermal Conductivity of CE 16xl6 Fuel Assemblies.............................................................................................................. A.4.9-1 A.4.9.1 Transverse Thermal Conductivity of Fuel in Helium and Vacuum.............................................................................................. A.4.9-1 A.4.9.2 Axial Thermal Conductivity ofFuel.................................................. A.4.9-3 Validation ofFLUENTTM/ICEPAK' Computer Codes against NUHOMS-7P Test Data...................................................................................... A.4.10-1 A.4.11 Supplemental Information......................................*.............................................. A.4.11-1 A.4.11.1 References........................................................................................ A.4.11-1 A.4.11.2 Thermal Analysis of Blocked Vent Accident Condition for NUHOMS Advanced Horizontal Storage Module using CFD Method............................................................................................. A.4.11-4 A.5 SHIELDING EVALUATION................................................................................................. A.5.1-1 A.5.1 Discussion and Results........... :................................................................................ A.5.1-3 A.5.2 Source Specification................................................................................................ A.5.2-1 A.5.2.1 Gamma Source................................................................................... A.5.2-4 A.5.2.2 Neutron Source Term......................................................................... A.5.2-4 A.5.2.3 Response Functions.for Alternate Nuclear Parameters...................... A.5.2-4 A.5.3 Model Specification................................................................................................ A.5.3-1 A.5.3.1 Description of the Radial and Axial Shielding Configurations.......... A.5.3-1 A.5.3.2 Shield Regional Densities.................................................................. A.5.3-3 A.5.4 Shielding Evaluation............................................................................................... A.5.4-1 A.5.4.1 Computer Program............................................................................. A.5.4-1 A.5.(2 Flux-to-Dose Rate Conversion........................................................... A.5.4-2 A.5.5 Supplemental Information....................................................................................... A.5.5-1 A.5.5.1 References.......................................................................................... A.5.5-1 A.5.5.2 Sample SAS2H Input Listing............................................................ A.5.5-3 A.5.5.3 Sample ANISN Model (Neutron Response Function for AHSM).............................................................................................. A.5.5-5 A.5.5.4 Sample AHSM MCNP Analysis Input Files..................................... A.5.5-9 A.5.5.5 Sample OS197H MCNP Analysis Input Files......... :...................... A.5.5-79 A.6 CRITICALITYEVALUATION................................................................................................ A.6-1 A.6.1 Discussion and Results............................................................................................ A.6.1-1 A.6.2 Spent Fuel Loading................................................................................................. A.6.2-1 A.6.3 Model Specification................................................................................................ A.6.3-1 ANUH-01.0150 iii
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9. 03/19 I A.6.3.l Description of Criticality Analysis Model......................................... A.6.3-1 A.6.3.2 Neutron Absorber Panel Material Efficacy........................................ A.6.3-5 A.6.4 Criticality Calculation............................................................................................. A.6.4-1 A.6.4.1 Calculational Method......................................................................... A.6.4-1 A.6.4.2 Normal Operating Conditions (NOC)................................................ A.6.4-1 A.6.4.3 Hypothetical Accident Conditions (HAC)......................................... A.6.4-4 A.6.4.4 Damaged Fuel Models....................................................................... A.6.4-4 A.6.4.5 Fuel Assembly Replacement............................................................ A.6.4-10 A.6.4.6 Assembly Reconstitution................................................................. A.6.4-10 A.6.4.7 Criticality Results.................................,........................................... A.6.4-10 A.6.4.8 Summary and Conclusions............................................................... A.6.4-12 A.6.5 Critical Benchmark Experiments............................................................................ A.6.5-1 A.6.5.1 Benchmark Experiments and Applicability....................................... A.6.5-1 A.6.5.2 Results of the Benchmark Calculations............................................. A.6.5-2 A.6.6 Supplemental Information....................................................................................... A.6.6-1 A.6.6.1 References.......................................................................................... A.6.6-1 A.6.6.2 KENO V.a Input Files........................................................................ A.6.6-2 A.7 CONFINEMENT.............................................. *......................................,............................... A.7.1-1 A. 7.1 Confinement Boundary........................................................................................... A. 7.1-1 A.7.1.1 Confinement Vessel................................................................. :......... A.7.1-1 A. 7.1.2 Confinement Penetrations.................................................................. A. 7.1-2 A.7.1.3 Seals and Welds................................................................................. A.7.1-2 A.7.1.4 Closure............................................................................................... A.7.1-2 A.7.1.5 Leak Testing Requirements............................................................... A.7.I'-2 A.7.2 Requirements for Normal Conditions of Storage.................................................... A.7.2-1 A.7.2.l Release of Radioactive Material........................................................ A.7.2-1 A. 7.2.2 Pressurization of Confinement Vessel............................................... A. 7.2-1 A. 7.3 Confinement Requirements for Hypothetical Accident Conditions........................ A. 7.3-1 A.7.3.1 Fission Gas Products.......................................................................... A.7.3-1 A.7.3.2 Release of Contents............................................................................ A.7.3-1 A.7.4 Supplemental Data.................................................................................................. A.7.4-1 A.7.4.1 Confinement Monitoring Capability.................................................. A.7.4-1 A.7.4.2 References..... ~.................................................................................... A.7.4-1 A.8 OPERA TING PROCEDURES................................................................................................ A.8.1-1 A.8.1 Procedures for Loading the 24PT4-DSC and Transfer to the AHSM..................... A.8.1-1 A.8.1.1 Narrative Description......................................................................... A.8.1-1 A.8.2 Procedures for Unloading the 24PT4-DSC................... :......................................... A.8.2-1 A.8.2.1 24PT4-DSC Retrieval from the AHSM............................................. A.8.2-1 A.8.2.2 Removal of Fuel from the 24PT4-DSC............................................. A.8.2-1 A.8.3 Supplemental Information....................................................................................... A.8.3-1 A.8.3.1 Other Operating Systems................................................................... A.8.3-1 A.8.3.2 Operation Support System................................................................. A.8.3-1 A.8.3.3 Control Room and/or Control Areas............................,..................... A.8.3-1 A.8.3.4 Analytical Sampling........................................................................... A.8.3-1 A.8.3.5 References.......................................................................................... A.8.3-1 ANUH-01.0150 iv
Advanced NUHOMS System Updated Final. Safety Analysis Report Rev. 9. 03/19 I A.9 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM.............................................. A.9.1-1 A.9.1 Acceptance Criteria................................................................................................. A.9.1-1 A.9.1.1 Visual Inspection................................................................................ A.9.1-1 A.9.1.2 Structural............................................................................................ A.9.1-1 A.9.1.3 Leak Tests and Hydrostatic Pressure Tests........................................ A.9.1-1 A.9.1.4 Components....................................................................................... A.9.1-1 A.9.1.5 Shielding Integrity.............................................................................. A.9.1-2 A.9.1.6 Thermal Acceptance.......................................................................... A.9.1-2 A.9.1. 7 Neutron Absorber Tests..................................................................... A.9.1-2 A.9.1.8 Deleted............................................................................................... A.9.1-3 A.9.1.9 Tests for B4C Encapsulated in Stainless Steel Tubes......................... A.9.1-3 A.9.2 Pre-Operational Testing and Maintenance Program............................. :................. A.9.2-1 A.9.2.1 Subsystems Maintenance................................................................... A.9.2-1 A.9.2.2 Valves, Rupture Discs, and Gaskets on Confinement Vessel.......... A.9.2-la A.9.3 Training Program...................................................................................... :.. :.......... A.9.3-1 A.9.3.1 Program Description.......................................................................... A.9.3-1 A.9.3.2 Retraining Program............................................................................ A.9.3-1 A.9.3.3 Administration and Records............................................................... A.9.3-1 A.9.4 Supplemental Information....................................................................................... A.9.4-1 A.9.4.1 References........................................... :.............................................. A.9.4-1 A.10 RADIATION PROTECTION............................................................................................... A.10.1-1 A. l 0.1 Ensuring that Occupational Radiation Exposures Are as Low as Reasonably Achievable (ALARA)........................................................................ A.10.1-1 A. l 0.2 Radiation Protection Design Features.......................................................... :........ A.10.2-1 A.10,2.1 Advanced NUHOMS System DesignFeatures.............................. A.10.2-1 A.10.2.2 Radiation Dose Rates....................................................................... A.10.2-1 A.10.2.3 AHSM Dose Rates........................................................................... A.10.2-4 A.10.2.4 ISFSI Array....................................... :.............................................. A.10.2-5 A.10.3 Estimated Onsite and Offsite Dose Assessment.................................................... A.10.3-1 A.10.3.1 Occupational Exposures................................................................... A.10.3-1 A.10.3.2 Public Exposure............................................................................... A.10.3-3 A.10.4 Supplemental Information..................................................................................... A.10.4-1 A.10.4.1 References........................................................................................ A.10.4-1 A.11 ACCIDENT ANALYSES........................................................................................................ A.11-1 A.11.1 Off-Normal Operations......................................................................................... A.11.1-1 A.11.1.1 Off-Normal Transfer Loads............................................................. A.11.1-1 A.11.1.2 Extreme Ambient Temperatures...................................................... A.11.1-1 A.11.1.3 Radiological Impact from Off-Normal Operations.......................... A.11.1-2 A.11.2 Postulated Accidents............................................................................................. A.11.2-1 A.11.2.1 Earthquake....................................................................................... A.11.2-1 A.11.2.2 Tornado Wind Pressure and Tornado Missiles................................ A.11.2-1 A.11.2.3 Flood................................................................................................ A.11.2-1 A.11.2.4 Fire/Explosion.................................................................................. A.11.2-2 A.11.2.5 Accidental Drop of the 24PT4-DSC Inside the Transfer Cask........ A.11.2-2 ANUH-01.0150 V
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9. 03/19 j A.13 A.14 A.11.3 A.12 A.12.1 A.12.2 A.12.3 A.11.2.6 Lightning.......................................................................................... A.11.2-3 A.11.2. 7 Blockage of Air Inlet and Outlet Openings...................................... A.11.2-3 A.11.2.8 Accidental Pressurization of the 24PT4-DSC.................................. A.l l.2-3 A.11.2.9 Burial................................................................................................ A.11.2-4 A.11.2.10 Inadvertent Loading of a Newly Discharged Fuel Assembly.......... A.11.2-4 Supplemental Information..................................................................................... A.11.3-1 A.11.3.1 References........................................................................................ A.11.3-1 CONDITIONS FOR CASK USE: OPERA TING CONTROLS AND LIMITS OR TECHNICAL SPECIFICATIONS AND BASIS FOR TECHNICAL SPECIFICATIONS FOR24PT4 SYSTEM..................................... A.12-1 TECHNICAL SPECIFICATIONS.......................................................................... A.12-1 Functional and Operating Limits.............................................................................. A.12-2 Limiting Condition for Operation (LCO) and Surveillance Requirements (SR)
Applicability.............................................................................................................. A.12-4 A.12.3.1 DSC Integrity.................................................................................... A.12-11 QUALITY ASSURANCE........................................................................................................ A.13-1 DECOMMISSIONING......................................................................................................... A.14.1-1 A.14.1 Decommissioning Considerations......................................................................... A.14.1-1 A.14.2 Supplemental Informational.................................................................................. A.14.2-1 A.14.2.1 References.. :..................................................................................... A.14.2-1
- ANUH-01.0150 vi
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9. 03/19 I Table A.1.2-1 Table A.2.1-1 Table A.2.1-2 Table A.2.1-3 Table A.2.1-4 Table A.2.1-5 Table A.2.1-6 Table A.2.1-7 Table A.2.1-8 Table A.2.1-9 Table A.2.1-10 Table A.2.1-11 Table A.2.1-12 Table A.2.5-1 Table A.3.1-1 Table A.3.1-2 Table A.3.1-3 Table A.3.1-4 Table A.3.1-5 Table A.3.1-6 Table A.3.2-1 Table A.3.3-1 Table A.3.3-2 Table A.3.3-3 ANUH-01.0150 LIST OFT ABLES Key Design Parameters of the Advanced NUHOMS System Components........................................................................................................ A.1.2-4 PWR Fuel Specification oflntact Fuel to be Stored in NUHOMS 24PT4-DSC............................... :................ :....................................................... A.2.1-4 PWR Fuel Specifications of Damaged Fuel to be Stored in NUHOMS 24PT4-DSC........................................................................................................ A.2.1-5 PWR Fuel Assembly Design Characteristics..................................................... A.2.1-6 Maximum Fuel Enrichment v/s Neutron Poison Requirements for 24PT4-DSC........................................................................................................ A.2.1-7 PWR Fuel Qualification Table for 1.26 kW per Assembly for the NUHOMS 24PT4-DSC.................................................................................... A.2.1-8 PWR Fuel Qualification Table for 1.2 kW per Assembly for the NUHOMS 24PT4-DSC.................................................................................... A.2.l-9 PWR Fuel Qualification Table for 1.0 kW per Assembly for the NUHOMS 24PT4-DSC.................................................................................. A.2.l-10 PWR Fuel Qualification Table for 0.9 kW per Assembly for the NUHOMS 24PT4-DSC......................................................... :........................ A.2.1-11 PWR Fuel Qualification Table for 1.26 kW per Assembly for the NUHOMS 24PT4-DSC, Reconstituted Fuel with Stainless Steel Rods......... A.2.1-12 PWR Fuel Qualification Table for 1.2 kW per Assembly fo~ the NUHOMS 24PT4-DSC, Reconstituted Fuel with Stainless Steel Rods......... A.2.1-13 PWR Fuel Qualification Table for 1.0 kW per Assembly for the NUHOMS 24PT4-DSC, Reconstituted Fuel with Stainless Steel Rods......... A.2.1-14 PWR Fuel Qualification Table for 0.9 kW per Assembly for the NUHOMS 24PT4-DSC, Reconstituted Fuel with Stainless Steel Rods......... A.2.1-15 Advanced NUHOMS System Major Components and Safety Classification...................................................................................................... A.2.5-3 Codes and Standards for the Design, Fabrication and Construction of 24PT4-DSC Principal Components.................................................................... A.3.1-6 Stress Criteria for Partial Penetration Pressure Boundary Welds....................... A.3.1-7 24PT4-DSC Load Combinations and Service Levels........................................ A.3.1-8 24PT4-DSC Internal Pressure Loads............................................................... A.3.1-10 Alternatives to the ASME Code for the 24PT4-DSC (NB).............................. A.3.1-11 Alternatives to the ASME Code for the 24PT4-DSC Basket (NG/NF)........... A.3.1-12 Weights and Centers of Gravity of the 24PT4-DSC.......................................... A.3.2-2 Static Mechanical Properties for ASTM B29 Lead............................................ A.3.3-3 ASME Code Material Properties for SA-533 Grade B Class 1 Carbon Steel.................................................................................................................... A.3.3-4 ASME Code Material Properties for SA-479 Type XM-19.............................. A.3.3-5 vii
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9. 03/19 I Table A.3.5-1 Table A.3.5-2 Table A.3.5-3 Table A.3.5-4 Table A.3.6-1 Table A.3.6-2 Table A.3.6-3 Table A.3.6-4 Table A.3.6-5 Table A.3.6-6 Table A.3.6-7 Table A.3.6-8 Table A.3.6-9 Table A.3.6-10 Table A.3.6-11 Table A.3.6-12 Table A.3.6-13 Table A.3.6-14 Table A.4.1-1 Table A.4.1-2 Table A.4.1-3 Table A.4.4-1 Table A.4.4-2 Table A.4.4-3 Table A.4.4-4 Table A.4.4-5 Table A.4.4-6 Table A.4.4-7 ANUH-01.0150 Summary of Fuel Assembly Thermal and Irradiation Growth Calculations........................................................................................................ A.3.5-4 Fuel Assembly Properties................................................................................... A.3.5-5 Material Properties for Fuel Cladding............................................ :................... A.3.5-6 Fuel Assembly Loads......................................................................................... A.3.5-7 24PT4-DSC On-Site Load Combinations........................................................ A.3.6-18 24PT4-DSC Shell Assembly Normal and Off-Normal Operating Condition Maximum Stress Intensities............................................................ A.3.6-20 24PT4-DSC Shell Assembly Accident Condition Maximum Stress Intensities......................................................................................................... A.3.6-21 24PT4-DSC Shell Assembly Results for Normal and Off-Normal Load Combinations................................................................................................... A.3.6-22 24PT4-DSC Shell Assembly Results for Accident Level C Load Combinations................................................................................................... A.3.6-23 24PT4-DSC Shell Assembly Results for Accident Level D Load Combinations................................................................................................... A.3.6-24 Summary of Spacer Disc Maximum Stress Ratios........................................... A.3.6-25 Summary of Guidesleeve Assembly Maximum Stress Ratios......................... A.3.6-26 Summary of Results for Support Rod Assemblies........................................... A.3.6-27 Maximum/Minimum Forces/Moments in the Rail Components in the Local System.................................................................................................. A.3.6-27a Maximum/Minimum Forces/Moments in the Rail Extension Plates in the Local System............................................................................................ A.3.6-27b Maximum/Minimum Axial Forces in the Cross Member Components......... A.3.6-27c Rail Component, Results of Evaluation......................................................... A.3.6-27d Extension Plates and Cross Members, Results of Evaluation........................ A.3.6-27e Component Minimum and Maximum Temperatures in the Advanced NUHOMS System (Storage or Transfer Mode) for Normal Conditions.......... A.4.1-3 Component Minimum and Maximum Temperatures in the Advanced NUHOMS System (Storage or Transfer Mode) for Off-Normal Conditions....................... :.................................................................................. A.4.1-4 Component Minimum and Maximum Temperatures in the Advanced NUHOMS System (Storage and Transfer) for Accident Conditions............... A.4.1-5 AHSM Bulk Air Temperatures........................................................................ A.4.4-13 AHSM Thermal Analysis Results Summary.................................................... A.4.4-14
- AHSM Peak Component Temperatures at Normal/Off-Normal Conditions............................................................................ :............................ A.4.4-15 24PT4-DSC Maximum Shell Temperatures for Normal/Off-Normal Conditions........................................................................................................ A.4.4-16 24PT4-DSC Maximum Shell Temperatures for Accident Conditions............. A.4.4-17 24PT4-DSC Basket Temperature Results........................................................ A.4.4-18 24PT4-DSC Maximum Fuel Cladding Temperature Results........................... A.4.4-19 viii
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9. 03/19 I Table A.4.4-8 Table A.4.4-9 Table A.4.4-10 Table A.4.4-11 Table A.4.7-1 Table A.4.7-2 Table A.4.7-3 Table A.4.9-1 Table A.4.9-2 Table A.4.9-3 Table A.4.9.-4 Table A.4.9-5 Table A.4.10-1 Table A.4.11.2-1 Table A.4.11.2-2 Table A.4.11.2-3 Table A.4.11.2-4 Table A.4.11.2-5 Table A.5.1-1 Table A.5.1-2 Table A.5.1-3 Table A.5.1-4 Table A.5.1-5 Table A.5.2-1 Table A.5.2-2 Table A.5.2-3 Table A.5.2-4 Table A.5.2-5 Table A.5.2-6 Table A.5.2-7 Table A.5.2-8 Table A.5.2-9 Table A.5.2-10 Table A.5.3-1 Table A.5.3-2 Table A.5.3-3 Table A.5.4-1 Table A.6.1-1 Table A.6.2-1 Table A.6.3-1 ANUH-01.0150 Summary of Cases Considered for Thermal Stress Analysis........................... A.4.4-20 Fuel Assembly Characteristics for Pressure Analysis...................................... A.4.4-21 24PT4-DSC Cavity Pressure Analysis Summary............................................. A.4.4-22 Technical Specifications 5.2.5.b Temperature Monitoring Limits for the 24PT4 DSC....................................................................................................... A.4.4-23 Gaps between Components of ANSYS Model at the Spacer Disc Plane........... A.4.7-7 Vacuum Drying Results following Blowdown with Air or Helium................... A.4.7-8 Summary of Water Heatup Calculation............................................................. A.4.7-9 Summary of Design Data For CE 16x16 Fuel Assembly Type......................... A.4.9-4 Thermal Properties Used in Calculation of Fuel Effective Conductivities........ A.4.9-5 CE 16x16 Computed Effective Transverse Thermal Conductivity.................... A.4.9-6 CE 16x16 Fuel Assembly Transverse Thermal Conductivities.......................... A.4.9-7 CE 16x16 Axial Thermal Conductivity.............................................................. A.4.9-7 Test Vs. Predicted Temperatures for 7P DSC in HSM.................................... A.4.10-5 AHSM lnsolation............................................................................................ A.4.11-13 Design Load Cases......................................................................................... A. 4.11-14 Summary of Convergence ofCFD Models..................................................... A.4.11-15 Maximum Component Temperatures for Off-Normal Condition................... A.4.11-16 Maximum Component Temperatures for Blocked Vent Accident Condition.. A. 4.11-17 Advanced NUHOMS System Shielding Materials.......................................... A.5.1-4 Summary of AHSM Dose Rates........................................................................ A.5.1-5 Transfer Cask (Loading/Unloading/Transfer Operations) Side Dose Rate Summary............................................................................... _............................. A.5.1-6 Transfer Cask (Loading/Unloading/Transfer Operations) Top End Dose Rate Summary......,............................................................................................. A.5.1-7 Transfer Cask (Transfer Operations) Bottom End Dose Rate Summary........... A.5.1-7 Fuel Assembly Region Materials, Masses, and Lengths.................................... A.5.2-7 Elemental Composition ofL WR Fuel-Assembly Structural Materials*.............. A.5.2-8 Flux Fraction By Assembly Region................................................................... A.5.2-9 CASK-81 Energy Group Structure................................................................... A.5.2-10 Design Basis Gamma Sources (per assembly)................................................. A.5.2-11 Design Basis Neutron Source (per assembly).................................................. A.5.2-12 AHSM and TC "Response Function" for Evaluating Fuel with Alternate Parameters......................................................................................................... A.5.2-13 "Response Function" Evaluation of Design Basis Source Terms.................... A.5.2-14 "Response Function" Evaluation of Candidate Fuel Assembly Source Terms 57 GWd/MTU, 3.8 wt.% U-235, 8-year Cooled Fuel Case............................ A.5.2-15 Relative Contribution of Source Terms to Dose Rates.................................... A.5.2-16 Materials Composition and Atom Number Densities (Dry)............................... A.5.3-4 Materials Composition and Atom Densities During Decontamination and Wet Welding Stage Calculation......................................................................... A.5.3-6 Flux to Dose Rate Conversion Factors............................................................... A.5.3-8 Normalized Bum-Up Shape for CE 16x16 Fuel Assembly............................... A.5.4-3 Summary of Limiting Criticality Evaluations for the CE 16x16 Fuel Assembly............................................................................................................ A.6.1-2 Fuel Parameters for Criticality Analysis of the CE 16x16 Fuel Assemblies...... A.6.2-1 Geometric Parameters Used in the Criticality Analysis..................................... A.6.3-7 ix
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9. 03/19 j Table A.6.3-2 Table A.6.3-3 Table A.6.3-4 Table A.6.4-1 Table A.6.4-2 Table A.6.4-3 Table A.6.4-4 Table A.6.4-5 Table A.6.4-6 Table A.6.4-7 Table A.6.4-8 Table A.6.4-9 Table A.6.4-10 Table A.6.4-11 Table A.6.4-12 Table A.6.4-13 Table A.6.4-14 Table A.6.4-15 Table A.6.4-16 Table A.6.4-17 Table A.6.5-1 Table A.6.5-2 Table A.6.5-3 Table A.10.2-1 Table A.10.2-2 Table A.10.2-3 Table A.10.2-4 Table A.10.2-5 Table A.10.2-6 Table A. I 0.2-7 Table A.10.2-8 Table A. I 0.2-9 Table A.10.3-1 ANUH-01.0150 Parameters for Poison Material Used in Criticality Analysis........................... A.6.3-10 Comparison of Criticality Model vs. Drawing Parameters (DSC Parameters)....................................................................................................... A.6.3-11 Comparison of Criticality Model vs. Drawing Parameters (Bora!
Parameters).................................... :.................................................................. A.6.3-13 Parametric Study Results - B-10 Areal density of 0.025 g/cm2 (Type A Basket).............................................................................................................. A.6.4-15 NOC Moderator Varying Results - B-10 Areal density of 0.068 g/cm2 (Type B Basket)............................................................................................... A.6.4-16 NOC Moderator Varying Results - B-10 Areal density of 0.025 g/cm2 (Type A Basket)............................................................................................... A.6.4-17 RAC Moderator Varying Results - B-10 Areal density of 0.068 g/cm2 (Type B Basket)............................................................................................... A.6.4-18 RAC Moderator Varying Results - B-10 Areal density of 0.025 g/cm2 (Type A Basket).............................................................".................................. A.6.4-19 Damaged Fuel, Rod Pitch Varying Results - B-10 Areal density of 0.068 g/cm2 (Type B Basket)..................................................................................... A.6.4-20 Damaged Fuel, Rod Pitch Varying Results with Rod Addition or Subtraction....................................................................................................... A.6.4-21 Damaged Fuel, Single-ended Shear................................................................ A.6.4-22 Damaged Fuel, Double-ended Shear............................................................... A.6.4-22 Damaged Fuel, Rod Pitch Cases-External Moderator Density Varying........ A.6.4-23 Damaged Fuel, Single Shear Cases - External Moderator Density Varying............................................................................................................. A.6.4-23 Damaged Fuel, Double Shear Cases - External Moderator Density Varying............................................................................................................. A.6.4-24 Damaged Fuel, Bare Fuel Added..................................................................... A.6.4-24 Summary of Maximum Enrichment for the Damaged Fuel Assemblies.......... A.6.4-25 Empty Fuel Assembly Locations..................................................................... A.6.4-26 Reconstituted Fuel Assemblies........................................................................ A.6.4-26 Storage Requirements by Fuel Enrichment...................................................... A.6.4-26 Benchmarking Results..................................... :************************************************** A.6.5-3 USL-1 Results.................................................................................................... A.6.5-7 USL Determination for Criticality Analysis....................................................... A.6.5-8 MCNP Front Detector Dose Rate Results for a Single AHSM........................ A.10.2-6 MCNP Side Detector Dose Rate Results for a Single AHSM......................... A.I 0.2-7 MCNP Back Detector Dose Rate Results for a Single AHSM........................ A.10.2-8 MCNP Front Detector Dose Rate Results for a 2x10 ISFSI............................ A.10.2-9 MCNP Side Detector Dose Rate Results for a 2x10 ISFSI............................ A.10.2-10 AHSM Gamma-Ray Spectrum Calculation Results....................................... A.10.2-11 AHSM Neutron Spectrum Calculations......................................................... A.10.2-12 Summary of AHSM Surface Activities.......................................................... A.10.2-13 ANISN Model Details.................................................................................... A. I 0.2-13 Advanced NUHOMS System Operations Estimated Time for Occupational Dose Calculations...................................................................... A.I 0.3-4 X
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9. 03/19 I Figure A.I.I-I Figure A.1.2-1 Figure A.2.1-1 Figure A.2.1-2 Figure A.2.1-3 Figure A.2.1-4 Figure A.3.1-1 Figure A.3.1-2 Figure A.3.1-3 Figure A.3.1-4 Figure A.3.2-1 Figure A.3.2-2 Figure A.3.6-1 Figure A.3.6-2 Figure A.3.6-3 Figure A.3.6-4 Figure A.3.6-5 Figure A.3.6-6 Figure A.3.6-7 Figure A.4.4-1 Figure A.4.4-2 Figure A.4.4-3 Figure A.4.4-4 Figure A.4.4-5 Figure A.4.4-6 Figure A.4.4-7 Figure A.4.4-8 Figure A.4.4-9 Figure A.4.4-10 Figure A.4.4-11 Figure A.4.4-12 Figure A.4.4-13 ANUH-01.0150 LIST OF FIGURES Page Advanced NUHOMS System 24PT4-DSC...................................................... A.1.1-3 24PT4-DSC ASME Code Boundary.................................................................. A.1.2-5 24PT4-DSC Heat Load Configurations #1, kW/Assembly.............................. A.2.1-16 24PT4-DSC Heat Load Configurations #2, kW/Assembly.............................. A.2.1-17 24PT4-DSC Heat Load Configurations #3, kW/Assembly.............................. A.2.1-18 Location of Failed Fuel Cans Inside 24PT4-DSC......................... :.................. A.2.l-I9 Advanced NUHOMS System 24PT4-DSC Canister Shell Assembly............ A.3.1-13 Advanced NUHOMS System 24PT4-DSC Pressure Boundary Location............................................................................................................ A.3.1-14 Advanced NUHOMS System 24PT4-DSC Canister Basket (Side View)................................................................................................................ A.3.1-15 Advanced NUHOMS System 24PT4-DSC Canister Basket & Shell (Side and Top End View)................................................................................. A.3.1-16 Schematic Location of Center of Gravity of the 24PT4-DSC............................ A.3.2-3 Schematic Location of Center of Gravity of the 24PT4-DSC in the AHSM.................... ~........................................................................................... A.3.2-4 24PT4-DSC Shell Assembly Axisymmetric Analysis Analytical Model........ A.3.6-28 24PT4-DSC Shell Assembly 3D ANSYS Models........................................... A.3.6-29 24PT4-DSC Load Support for Shell and Spacer Disc Analyses...................... A.3.6-30 Typical 24PT4-DSC Spacer Disc ANSYS Model for In-Plane Loads (Half Symmetry)................................................................................. :............ A.3.6-31 Typical 24PT4-DSC Spacer Disc ANSYS Model for In-Plane Loads (Full Symmetry)............................................................................................... A.3.6-32 Typical 24PT4-DSC Spacer Disc ANSYS Model for Out-of-Plane Loads (Quarter Symmetry)............................................................................... A.3.6-33 Load Application to 24PT4-DSC Spacer Disc................................................. A.3.6-34 Isometric, Wireframe View of the AHSM Model Layout................................ A.4.4-24 Isometric View of Seven (7) Axial Segments Used to Simulate DSC within Module... :.............................................................................................. A.4.4-25 Isometric View of 24PT4-DSC & Heat Shield Layout within Model............. A.4.4-26 Elevation View of Meshing at Z-Y Plane of AHSM....................................... A.4.4-27 Temperature Distribution on 24PT4-DSC Surface.......................................... A.4.4-28 Temperature Distribution on AHSM Heat Shield Surfaces............................. A.4.4-29 Velocity Profile along Y-Z Plane at Center of AHSM..................................... A.4.4-30 Isometric, Wire frame View of Model Layout.................................................. A.4.4-31 24PT4-DSC Spacer Disc Schematic................................................................ A.4.4-32 Plan View of Meshing at 24PT4-DSC Spacer Disc......................................... A.4.4-33 24PT4-DSC HLZC #1, kW/Assembly............................................................. A.4.4-34 24PT4-DSC HLZC #2, kW/Assembly............................................................. A.4.4-35.
24PT4-DSC HLZC #3, kW/Assembly............................................................. A.4.4-36 xi
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9. 03/19 j Figure A.4.4-14 Figure A.4.4-15 Figure A.4.4-16 Figure A.4.4-17 Figure A.4.4-18 Figure A.4.4-19 Figure A.4.4-20 Figure A.4.4-21 Figure A.4.4-22 Figure A.4.4-23 Figure A.4.4-24 Figure A.4.4-25 Figure A.4.4-26 Figure A.4.4-27 Figure A.4.4-28 Figure A.4.6-1 Figure A.4.6-2 Figure A.4.6-3 Figure A.4.7-1 Figure A.4.7-2 Figure A.4.7-3 Figure A.4.7-4 Figure A.4.7-5 Figure A.4.7-6 Figure A.4.7-7 Figure A.4.7-8 Figure A.4.7-9 Figure A.4.9-1 Figure A.4.9-2 ANUH-01.0150 24PT4-DSC Shell Temperatures, Bounding Condition in AHSM................... A.4.4-37 Fuel Cladding Temperature Distribution within 24PT4-DSC Basket, Bounding Condition in AHSM......................................................................... A.4.4-38 24PT4-DSC Spacer Disc Temperature Distribution, Bounding Condition in AHSM......................................................................................... A.4.4-39 24PT4-DSC Poison Sheet Temperature Distribution, Bounding Condition in AHSM......................................................................................... A.4.4-40 Velocity Distribution within 24PT4-DSC Basket, Bounding Condition in AHSM.......................................................................................................... A.4.4-41 24PT4-DSC Shell Temperatures, Bounding Condition in TC......................... A.4.4-42 Fuel Cladding Temperature Distribution within 24PT4-DSC Basket, Bounding Condition in TC............................................................................... A.4.4-43 24PT4-DSC Spacer Disc Temperature Distribution, Bounding Condition in TC................................................................................................ A.4.4-44 24PT4-DSC Poison Sheet Temperature Distribution, Bounding Condition in TC............................................................................................. *... A.4.4-45 Velocity Distribution within 24PT4-DSC Basket, Bounding Condition in TC................................................................................................................. A.4.4-46
- Spacer Disc Radial Temperature Distribution (Storage Conditions)............... A.4.4-47 Spacer Disc Radial Temperature Distribution (Transfer Conditions, Hot)...... A.4.4-48 Spacer Disc Radial Temperature Distribution-40°F TC.................................. A.4.4-49 Temperature along Vertical Lines through Basket, Bounding Condition in AHSM, HLZC #l......................................................................................... A.4.4-50 Temperature along Vertical Lines through Basket, Bounding Condition in TC................................................................................................................. A.4.4-51 Transient Temperatures of24PT4-DSC Components during Blocked Vent Case-HLZC #1......................................................................................... A.4.6-6 Transient Temperatures of24PT4-DSC Components during Blocked Vent Case-HLZC #3.......................................................................................... A.4.6-6 OS197H Cask and 24PT4-DSC Response to Fire Accident.............................. A.4.6-7 Simplified Axial View of the 24PT4-DSC Basket Model............................... A.4.7-10 24PT4-DSC ANSYS Thermal Model; Front And Side Views........................ A.4.7-11 24PT4-DSC ANSYS Thermal Model, Spacer Disc......................................... A.4.7-12 24PT4-DSC ANSYS Thermal Model, Shell and Guidesleeve Assembly........ A.4.7-13 24PT4-DSC ANSYS Thermal Model, Fuel Assemblies.................................. A.4.7-14 Surface Elements for Radiation View Factor Calculation................................ A.4. 7-15 Gaps between Components of ANSYS Model................................................ A.4.7-16 Maximum Fuel Cladding Temperature during Vacuum Drying Using Air for Blowdown............................................................................................ A.4.7-17 Time to Reach Boiling Conditions inside 24PT4-DSC Cavity........................ A.4.7-18 Perspective View of CE 16x16 Thermal Model (1/4 Segment)......................... A.4.9-8 Finite Element Modeling of 1/4 Segment CE 16x16 Assembly........................ A.4.9-8 xii
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9. 03/19 I Figure A.4.9-3 Figure A.4.9-4 Figure A.4.10-1 Figure A.4.10-2 Figure A.4.10-3 Figure A.4.10-4 Figure A.4.10-5 Figure A.4.10-6 Figure A.4.10-7 Figure A.4.10-8 Figure A.4.10-9 Figure A.4.10-10 Figure A.4.10-11 Figure A.4.10-12 Figure A.4.10-13 Figure A.4.10-14 Figure A.4.11.2-1 Figure A.4.11.2-2 Figure A.4.11.2-3 Figure A.4.11.2-4 Figure A.4.11.2-5 Figure A.4.11.2-6 Figure A.4.11.2-7 Figure A.4.11.2-8 Figure A.5.1-1 Figure A.5.1-2 Figure A.5.1-3 Figure A.5.1-4 Figure A.5.2-1 Figure A.5.2-2 Figure A.5.4-1 Figure A.5.4-2 Figure A.5.4-3 Figure A.5.4-4 Figure A.5.4-5 Figure A.5.4-6 Figure A.5.4-7 Figure A.5.4-8 Figure A.5.4-9 ANUH-01.0150 Effective Transverse Thermal Conductivity for 1.0 kW CE 16x16 Fuel Assembly - Helium Filled Condition................................................................. A.4.9-9 Effective Transverse Thermal Conductivity for 1.0 kW CE 16x16 Fuel Assembly - Air Filled Condition....................................................................... A.4.9-9 Layout ofNUHOMS -7P HSM Array for Performance Testmg..................... A.4.10-7 General Geometry ofNUHOMS-7P Horizontal Storage Module.................. A.4.10-8 Thermocouple Location in Center Module (HSM-2)....................................... A.4.10-9 Thermocouple Locations 5'-2" from Front ofModules................................. A.4.10-10 Thermocouple Locations 7'-2" from Front of Modules................................. A.4.10-11 Thermocouple Locations 9'-3" from Front of Modules (Center)................... A.4.10-12 Thermocouple Locations 15'-6" from Front ofModules............................... A.4.10-13 Isometric, Wireframe View ofNUHOMS-7P HSM Model Layout............. A.4.10-14 Elevation View of Meshing along Axial Center Plane of 7P HSM............... A.4.10-15 Temperature Distribution on 7P DSC Surface............................................... A.4.10-16 Temperature Distribution on Heat Shield Surfaces........................................ A.4.10-17 Temperature Distribution on Side & Rear Concrete Surfaces....................... A.4.10-18 Axial Velocity Profile............................................................... :.................... A.4.10-19 Velocity Profile along X-Y Plane at Center of Module................................. A.4.10-20 CAD Model of AHSMwith 24PT4 DSC......................................................... A.4.11-18 Axial View of AHSM Mesh with 24PT4 DSC on the Symmetrical Mid-plane....................................................................... :....................................... A.4.11-19 Cross-Sectional View of the Half-Symmetrical AHSM Mesh with 24PT4 DSC................................................................................................................ A.4.11-20 Temperature Monitor of Various Components for Steady-State Off-Normal Simulation......................................................................................... A.4.11-21 Temperature Contour for AHSM Loaded with 24PT4 DSC at Steady-State Off-Normal Conditions.......................................................................... A.4.11-22 Temperature Rise of As-Built Thermocouple................................................. A.4.11-23 Temperature Monitor of Various Components for Off-Normal Accident Simulation....................................................................................................... A.4.11-23 Temperature Contour for AHSM Loaded with 24PT4 DSC for Blocked Vent Accident Condition at 25 Hours............................................................. A.4.11-24 Advanced NUHOMS System (24PT4-DSC in AHSM) Shielding Configuration..................................................................................................... A.5.1-8 24PT4-DSC Shielding Configuration................................................................ A.5.1-9 Right Elevation Cross Section View of AHSM............................................... A.5.1-10 Shielding Configuration of the TC................................................................... A.5.1-11 ANISN AHSM Model...................................................................................... A.5.2-17 ANISN TC Model............................................................................................ A.5.2-18 AHSM Bottom MCNP Model, (x,z) Cut........................................................... A.5.4-4 AHSM Bottom MCNP Model, (y,z) Cut........................................................... A.5.4-5 AHSM Top MCNP Model, (x,z) Cut................................................................. A.5.4-6 AHSM Top MCNP Model, (y,z) Cut................................................................. A.5.4-7 OS197H MCNP Model...................................................................................... A.5.4-8 OS197H Cask MCNP Model-Top Section...................................................... A.5.4-9 OS197H Cask MCNP Model-Bottom Section.............................................. A.5.4-10 OS197H Cask MCNP Model (Top) during Decontamination......................... A.5.4-11 OS197H Cask MCNP Model (Top) during Wet Welding............................... A.5.4-12 xiii
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9. 03/19 I Figure A.5.4-10 Figure A.6.3-1 Figure A.6.3-2 Figure A.6.3-3 Figure A.6.3-4 Figure A.6.3-5 Figure A.6.4-1 Figure A.6.4-2 ANUH-01.0150 OS 197H Cask MCNP Model (Top) during Dry Welding................................ A.5.4-13 KENO V.aModel of the 24PT4-DSC Basket.................................................. A.6.3-14 Exploded View of KENO V.a Model.............................................................. A.6.3-15 Structure of KENO V.a Model-UNIT 33...................................................... A.6.3-16 Structure of KENO V.a Model-UNIT 34...................................................... A.6.3-17 Cross Section of the CE 16x16 Fuel Assembly................................................ A.6.3-18 Fuel Assembly Cross Section Showing the Burnable Absorber Rod Configuration................................................................................................... A.6.4-27 Fuel Assemblies Located in the Inner Guidesleeve Corner Closest to the DSC Centerline (Assembly in Case)...................... :......................................... A.6.4-29 xiv
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9. 03/19 I Figure A.6.4-3 Figure A.6.4-4 Figure A.6.4-5 Figure A.6.4-6 Figure A.6.4-7 Figure A.6.4-8 Figure A.6.4-9 Figure A.6.4-10 Figure A.6.4-11 Figure A.6.4-12 Figure A.6.4-13 Figure A.7.1-1 Figure A.8.1-1 Figure A.8.2-1 Figure A.10.2-1 Figure A.10.2-2 ANUH-01.0150 Fuel Assemblies Moved Radially Outwards from the Center of the 24PT4-DSC (Assembly Out Case)................................................................... A.6.4-30 Fuel As~emblies Moved Towards the Upper Left Corner of Each Guidesleeve Assembly Upper Left Corner Case.............................................. A.6.4-31 Single-ended Shear Model............................................................................... A.6.4-32 Double-ended Shear Model.............................................................................. A.6.4-33 Geometry with Bare Fuel Rods Added............................................................ A.6.4-34 Loading Pattern for 4 Damaged Fuel Assemblies............................................ A.6.4-35 Loading Pattern for 12 Damaged Fuel Assemblies.......................................... A.6.4-36 Fuel Assembly with Guide Tubes and Poison Rodlets.................................... A.6.4-37 Example of 4 Empty Fuel Assembly Locations............................................... A.6.4-3 8 Example of a Reconstituted Fuel Assembly..................................................... A.6.4-39 Failed Fuel Cans Positions............................................................................... A.6.4-40 24PT4-DSC Confinement Boundary Welds...................................................... A.7.1-4 Advanced NUHOMS System Loading Operations Flow Chart....................... A.8.1-9 Advanced NUHOMS System Retrieval Operations Flow Chart...................... A.8.2-6 Annual Exposure from a Single AHSM as a Function of Distance............... A.10.2-14 Annual Exposure from a 2x10 AHSM Array as a Function of Distance....... A.10.2-15 xv
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I water-filled cask annulus. Therefore, this case results in the bounding thermal growth for all operating conditions.
There is adequate space within the 24PT4-DSC cavity for thermal and irradiation growth of the fuel assemblies. The minimum calculated gap is given in Table A.3.5-1.
A.3.5.3 Fuel Rod Integrity During Drop Scenario Fuel assembly properties are provided in Table A.3.5-2; material properties are provided in Table A.3.5-3 and fuel assembly loads are identified in Table A.3.5-4 for the calculation of fuel rod stresses and critical buckling loads due to cask side and end drop incidents.
A.3.5.3.1 Methodology A.3.5.3.1.1 Drop The drop analysis methodology is the same as presented in Section 3.5.3.1 for both side and comer drops.
A.3.5.3.2 Results Using the geometric and material properties in Table A.3.5-2 through Table A.3.5-4 and the methodology in Section 3.5.3.1, the analysis of the Westinghouse-CENP, Combustion Engineering 16x16 Zircaloy-4 clad fuel assemblies for 75g side and 25g comer drops and the methodology described above gives the following results:
The side drop allowable g-load is calculated to be 154g which exceeds the postulated 75g side load. For the comer drop, the critical axial buckling load is calculated to be 61.2g which, when combined with the side drop component, results in an interaction ratio of 0.36. This provides a factor of safety greater than 2 against fuel rod failure in a comer drop.
A.3.5.4 Fuel Unloading For unloading operations during the time period when the spent fuel pool is available, the 24PT4-DSC will be filled with spent fuel pool water through the siphon port. During this filling operation, the 24PT4-DSC vent port is maintained open with effluents routed to the plant's off-gas monitoring system. The NUHOMS operating procedures recommend that the 24PT4-DSC cavity atmosphere be sampled before introducing any reflood water into the 24PT4-DSC cavity.
When the pool water is added to a 24PT4-DSC cavity containing hot fuel and basket components, some of the water will flash to steam causing internal cavity pressure to rise. This steam pressure is released through the vent port. The procedures also specify that the flow rate and temperatures of the reflood water be controlled to ensure that the internal pressure in the 24PT4-DSC cavity is maintained at less than or equal to 20 psig. The reflood for the 24PT4-DSC is considered as a Service Level D event. The 24PT4-DSC is also evaluated for a Service Level D pressure of 100 psig. Therefore, there is sufficient margin in the 24PT4-DSC internal pressure during the reflooding event to assure that the 24PT4-DSC will not be over pressurized.
ANUH-01.0150 A.3.5-2 All changes on this page are Amendment 4.
Updated Advanced NUHOMS Updated System Final Safety Analysis Report Rev. 9, 03/19 Figure A.4.4-6 illustrates the temperature profile of the heat shields within the AHSM. The variation in temperature from back to front on the heat shield reflects the computed distribution of airflow over the DSC.
Figure A.4.4-7 illustrates the flow profile along a y-z plane through the center of the module.
The figure depicts the expected regions of flow recirculation within the inlet duct and in the plenum below the DSC. The figure also shows that, despite interior obstructions, the airflow is relatively evenly distributed along the length of the DSC. A maximum flow velocity of approximately 5.6 feet per second is seen to occur in the vertical exhaust duct at the rear of the module.
Table A.4.4-1 and Table A.4.4-2 summarize the results of the AHSM thermal analysis performed. Table A.4.4-3 summarizes the peak temperatures for the AHSM components based on this model. Table A.4.4-4 summarizes the peak DSC shell surface temperatures as a function of its geometry (from bottom to top) for the normal and off-normal conditions.
A.4.4.2.4 Monitoring of AHSM Temperature AHSM temperature monitoring is provided to alert operators to a possible blocked vent condition. The temperature rise of the as-built dual thermocouple locations is obtained by CFD analysis using the ANSYS FLUENT code as described in Section A. 4.11. 2. The corresponding Technical Specifications temperature limits for this location are provided in Table A.4.4-11.
A.4.4.3 Thermal Analysis of 24PT4-DSC in the TC The thermal analysis of the 24PT4-DSC in the TC is also split into separate models for the 24PT4-DSC and TC. This allows for independent calculation of24PT4-DSC internal temperatures, using the 24PT4-DSC shell temperatures calculated in the TC model as input.
The purpose of the TC analysis is to determine the 24PT4-DSC shell temperatures to be used as boundary conditions in a subsequent 24PT4-DSC basket thermal analysis described in Section A.4.4.4. The thermal analysis of the TC with total heat load of 24 kW is presented in Section 4.4.3. The shell temperatures were provided for the 24PT1-DSC for the required range of ambient conditions with a 24 kW heat load. These shell temperatures are directly applicable for 24PT4-DSC since the shell outside diameter, wall thickness, and materials are the same for both designs. Since the thermal analysis of the TC is based on a homogenized DSC model, a small difference in basket dimensions between 24PT1-DSC and 24PT4-DSC will have a negligible affect on the results.
A.4.4.3.1 TC Model Description See Section 4.4.3.1.
ANUH-01.0150 A.4.4-5 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I Table A.4.4-11 Technical Specifications 5.2.5.b Temperature Monitoring Limits for the 24PT4 DSC Max Temp Rise Max Temp (0F)
(OF)
(in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />)
Dual Thermocouple (y = 60", x = +/-15", z = -11.25")
200 8.5<1>
- 1.
Based on a 24 kW DSC heat load, as noted in Technical Specification Section 5.2.5.b. at the "as-built" dual thermocouple locations provided in the AHSM roof.
ANUH-01.0150 A.4.4-23 All changes on this page are Amendment 4.
Updated Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 A.4.7 Thermal Evaluation for Loading/Unloading Conditions All individual fuel assembly transfer operations occur when the 24PT4-DSC is in the spent fuel pool. The fuel is always submerged in free-flowing pool water permitting heat dissipation.
After fuel loading is complete, the 24PT4-DSC is removed from the pool, drained, dried, and backfilled with helium.
The three bounding loading conditions evaluated are (1) the heatup of the 24PT4-DSC before the cavity can be backfilled with helium (i.e., prior to blowdown), (2) the vacuum drying transient, and (3) steady state temperatures subsequent to helium backfill. Transient thermal analyses are performed to predict the heatup time history for the 24PT4-DSC components during these events.
The unloading operation considered is the reflood of the 24PT4-DSC with water.
A.4.7.l Vacuum Drying Thermal Analysis Analyses were performed for the vacuum drying condition in order to ensure that the steady state fuel cladding and 24PT4-DSC structural component temperatures remain below the maximum allowable material limits shown in Table A.4.7-2. In addition, a transient analysis was performed to ensure the requirements defined by ISG-11 [A4.21] for short-term operations (including vacuum during and helium backfilling operating conditions) are satisfied. According to ISG-11, the maximum fuel cladding temperature cannot exceed T1sG limit= 400 °C (752 °F) and the temperature difference during the thermal cycling of the cladding cannot exceed /),, T1sG limit=
65 °C (117 °F).
During vacuum drying operation, water in the DSC cavity is forced out of the cavity (blowdown operation) before the start of vacuum drying. Two alternate options for the gas medium used for the water blowdown operation are evaluated.
In the first option, air is used as the gas medium to remove water and subsequent vacuum drying occurs with air environment in the DSC cavity. In the second option, helium is used as the medium to remove water and subsequent vacuum drying occurs with helium environment in the DSC cavity.
In the thermal analysis for the vacuum drying transient, either air or helium is used as the medium present in the DSC cavity during vacuum drying process. Details of the thermal analysis performed for these two alternate options are described in the following sections.
A.4.7.1.1 Analysis Model For the vacuum drying thermal analysis, a three-dimensional slice of the 24PT4-DSC basket assembly and fuel is modeled near. the center of the active fuel region using the ANSYS computer code. This case has little convection due to low pressure environment and therefore does not justify the use of a resource intensive CFD based code. The 3-D slice spans from center to center of two spacer discs to account for the radial effect of conduction through the spacer discs. Heat transfer effects along the axis of the 24PT4-DSC (third dimension) outside hottest section between two adjacent spacer disc mid-planes are conservatively neglected by applying ANUH-01.0150 A.4.7-1 All changes on this page are Amendment 4.
Updated Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I
[A4.31]
FLUENT', Version 6.1, FLUENT, Inc., Lebanon, NH, 2003.
[A4.32]
ICEPAK', Version 4.1, FLUENT, Inc., Lebanon, NH, 2003.
[A4.33]
"Characteristics of Spent Fuel, High Level Waste, And Other Radioactive Wastes Which May Require Long-Term Isolation," DOE/RW-0184, Volume 3 of 6, dated December 1987.
[A4.34]
"Domestic Light Water Reactor Fuel Design Evolution, Volume III," Nuclear Assurance Corporation, September 1981, DOE/ET/47912-3.
[A4.35]
NUREG/CR-0497, A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior, MA TPRO - Version 11 (Revision 2),
EG&G Idaho, Inc., TREE-1280, September 1981.
[A4.36]
SAND90-2406, Sanders, T. L., et al., A Method for Determining the Spent-Fuel Contribution to Transport Cask Containment Requirements, TTC-1019, UC-820, November 1992.
[ A4.3 7]
"Spent Nuclear Fuel Effective Thennal Conductivity Report", prepared TRW Environmental Safety Systems, Inc. for DOE Civilian Radioactive Waste Management System (CRWMS), Report BBA000000-01717-5705-00010, Rev. 0, July 1996.
[A4.38]
"SONGS Unit 2/3 Fuel Assembly Materials and Masses" SCE No. N-1020-162.
Transnuclear, Inc. No. SCE-23.0100-11.
[A4.39]
K. Minato, et. al., "Thermal Conductivities oflrradiated U02 and (U, Gd)02", Journal of Nuclear Materials, 300 (2002) 57-64.
[A4.40]
Ranchi, et. al., "Effect of Bum-up on the Thermal Conductivity of Uranium Dioxide up to 100,000 MWdt", Journal ofNuclear Materials, 327 (2004) 58-76.
[A4.41]
ANSYS FLUENT, Version 17.1, ANSYS, Inc.
[A4.42]
ANSYS ICEMCFD, Version 17.1, ANSYS, Inc.
[A4.43]
US. NRC, "Computational Fluid Dynamics Best Practice Guidelines for Dry Cask Applications," NUREG-2152, March 2013.
ANUH-01.0150 A.4.11-3 All changes on this page are Amendment 4.
Updated Advanced NUHOMffiY System Updated Final Safety Analysis Report Rev. 9, 03/19 A. 4.11. 2 Thermal Analysis of Blocked Vent Accident Condition for NUHOMs© Advanced Horizontal Storage Module using CFD Method AHSM Temperature monitoring is provided to alert operators of a possible blocked vent condition. This section predicts the temperature rise of the thermocouple at the as-built location (y=60 ",x=+/-15 '~z=-11.25 ) in the AHSM loaded with the 24PT4 DSC, for the blocked vent accident condition using the ANSYS FLUENT [A4.41} CFD code. The blocked vent accident condition is considered for a duration of 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />.
The ANSYS FLUENT CFD code was used in Section A.4.4.2 to determine the thermal performance of the AHSM under steady-state conditions and the HEATING7 code was used in Section A. 4. 6 to determine the transient behavior of the AHSM during the blocked vent accident condition. In this Section, the ANSYS FLUENT CFD code is used to determine the temperature rise of the thermocouple during the blocked vent accident condition. This approach eliminates uncertainties, such as the difference in the initial temperatures between the two codes at the start of the blocked vent accident condition.
A.4.11.2.1 Design Input The design of the 24PT4 DSC and AHSM are based on the geometry summarized in Chapters 1 and A. I and the AHSM drawings provided in Section 1.5.2. The basket assembly, including the fuel assemblies, is modeled as a homogenous basket with effective properties.
Ambient Operating Conditions As specified in Section 4.1.2, a 24-hour daily average temperature of 107 °F, corresponding to the maximum off-normal ambient temperature of 107 ° F, is used in this evaluation.
Solar Insolation The insolation values from 10 CFR Part 71 [A4. l 3 }, applied over the exterior surfaces of the AHSMwalls, are listed in Table A.4.11.2-1.
Design Load Cases To determine the temperature rise of the thermocouple for the as-built configuration of the AHSM loaded with the 24PT4 DSC, the load cases_ listed in Table A.4.11.2-2 are evaluated for off-normal and accident conditions.
Load case 1 (LC #1) evaluates the off-normal hot storage condition and provides the initial conditions for the transient blocked vent accident condition. The transient blocked vent accident condition (LC#2) is evaluated for a duration of25 hours.
A.4.11.2.2 Methodology To determine the temperature rise of the thermocouple for the as-built configuration of the AHSM loaded with the 24PT4 DSC, a three-dimensional (3D), half-symmetrical, CFD model in ANSYS FLUENT is developed.
ANUH-01.0150 A.4.11-4 All changes on this page are Amendment 4.
Updated Advanced NUHOMsID System Updated Final Safety Analysis Report Rev. 9, 03/19 SectionA.4.11.2.3.l describes the steps used to generate a computer aided design (CAD) model for the AHSM in ANSYS ICEM CFD [A4.42]. Section A.4.11.2.3.2 describes the mesh generation for the AHSM and the 24PT4 DSC in ANSYS ICEM CFD. To simulate the airflow within the cavity of the AHSM, the low-Reynolds k-t: turbulence model with.full buoyancy effects is used in this evaluation. In addition to the convection within the AHSM cavity, the model also includes:
(1) the heat conduction within the homogeneous basket and the AHSM, (2) radiation heat transfer among the DSC shell, heat shields, and the AHSM cavity surfaces, (3) solar insolation through the AHSM.front wall and roof, and (4) heat dissipation.from the AHSM and the vent outlet via convection and radiation to the ambient environment. The detailed configuration of the CFD model in ANSYS FLUENT is described in Section A.4.11.2.3.3.
A.4.11.2.3 Thermal Model A.4.11.2.3.1 CAD Model A JD half-symmetric CAD model of the AHSM with the 24PT4 DSC is developed using ANSYS ICEMCFD. The model includes the AHSM concrete walls, roof, andfloor, heat shields, 24PT4 DSC shell (along with the top and bottom combined end plates), support structures and outlet vent cover. The basket assembly with fuel assemblies is modeled as a homogeneous basket..
Figure A.4.11.2-1 presents an isometric view of the AHSMCAD model loaded with a 24PT4 DSC. To simplify the model, the various cover plates at the top and bottom of the DSC are combined as effective regions. The DSC dimensions are taken.from Section A.4.4.2.1.
A.4.11.2.3.2 Meshing The CAD model is imported into ANSYS ICEM-CFD for meshing. The multi-block method is used to create the conformal structured mesh in ICEM-CFD using the "Blocking" method. A fine mesh is created on all the areas with a space ratio of approximately 1.2 and cell jumps are controlled to generate a high quality mesh with smooth transitions. The first cell height near the DSC is maintained as 0. 05 inch to capture the temperature gradients near the DSC.
The total computational domain comprises 40,643,152 hexahedral elements. Figure A.4.11. 2-2 shows an axial view of the hexahedral mesh on the symmetrical mid-plane of the AHSM model and the DSC. Figure A.4.11.2-3 shows a cross-sectional view of the half-symmetrical hexahedral mesh for the AHSM and the DSC at the axial middle section of the DSC.
A.4.11.2.3.3 CFD Modeling The mesh is scaled to the appropriate International System of Units (S.I) units and is imported to ANSYS FLUENT. The following steps describe the setup of the CFD model in ANSYS FLUENT.
A. Defining Geometry and Domains After importing the volumetric mesh, the cell zone conditions are defined, selecting the appropriate medium (fluid/solid) and appropriate material bases in the components.
ANUH-01.0150 A.4.11-5 All changes on this page are Amendment 4.
Updated Advanced NUHOMSW System Updated Final Safety Analysis Report Rev. 9, 03/19 I B. Selecting Physical Sub-models Based on the discussion presented in Section 7.2 of [A4.43], the low-Reynolds k-& turbulence model with full buoyancy effects is recommended for the turbulence model. The default values for the model constants are determined from various experiments for fundamental turbulent flows and work well for a wide range of wall-bounded and free shear flows. Therefore, the default model constants are used in this low-Reynolds k-& turbulence model.
The discrete*ordinates (DO) model is used as the radiation model. The DO methodology supports symmetry boundary conditions and requires modest computational resources.
C. Material Properties The properties for each material used in this evaluation, with the exception of the homogenized basket, are listed in Section 4.2. The effective properties of the homogenized basket are calculated based on weight, volume and material of the components as discussed in Section 4.7.3.
For ease of modeling, the thermal property inputs for ANSYS FLUENT are all in SI units.
D. Specifying Operating. Initial and Boundary Conditions Operating Conditions:
The operating conditions are specified to include the effects of gravity and operating density.
The effect of gravity is modeled as an acceleration of-9.81 mls2 in the Y direction.
According to [A4.43}, the correct operating density should be the density evalu[Jted at the air inlet conditions of pressure and temperature. Therefore an operating density of air equal to 1.1213 kglm3, which is based on the inlet temperature of 107 °F and atmospheric pressure, is specified for buoyancy flows.
Load Cases:
Steady-State Case:
To obtain the initial conditions of the accident blocked vent condition, a steady state analysis for the off-normal hot storage condition (LC#l) is performed.
Transient Case:
For the transient accident blocked vent condition (LC #2), results from the steady-state off-normal hot condition (LC #1) provide the initial conditions.
ANUH-01.0150 A.4.11-6 All changes on this page are Amendment 4.
Updated Advanced NUHOMS" System Updated Final Safety Analysis Report Inlet/outlet Boundary Condition Steady-State Case:
Rev. 9, 03/19 I Inlet and outlet vent boundary conditions are specified at the air inlet and air outlet.
Pressure and temperature are set to be the ambient pressure and temperature. To model the effect of screens located at the inlet and outlet, a flow loss coefficient of 0.58 [A4.30] is assumed to account for the resistances due to the wire mesh.
The turbulence boundary conditions at both the inlet and outlets are specified by using the "intensity and hydraulic diameter method. " Based on the dimensions of the inlet and outlet vents, the hydraulic diameters are calculated as 0.5067 mfor the inlet and 0.5265 mfor the outlet.
Based on the discussion in the FLUENT Users Guide, Section 6.3.2.1.3 of [A4.41], the turbulence intensities at the inlet and outlet are calculated as 4% for the inlet and 5% for the outlet.
Transient Case:
For the transient accident blocked vent accident condition, insulated wall boundary conditions are applied at both the inlet and outlet.
Radiation Boundary Condition Two walls are automatically created at the faces shared between the fluid and solid domains when reading the mesh into FLUENT. Typically the walls have names as '*wall]" and "wall]: shadow" as shown below. The face adjacent to the fluid region is "walll: shadow" and the wall adjacent to the solid region is "walll. " For radiation boundary condition, any emissivity needs to be specified on the wall adjacent to the fluid region, i.e., "wall] :
shadow."
Fluid Typical Fluid-Solid Interface in ANSYS FLUENT Using the above approach, the adjacent regions are checked and the emissivities of the materials are specified to compute the radiation heat transfer.
Symmetry Since only one half of the AHSM and the homogenized basket are modeled, the AHSM is symmetrical about the vertical mid-plane parallel to the axial direction.
ANUH-0 1.0150 A.4.1 1-7 A ll changes on this page are Amendment 4.
Updated Advanced NUHOMsij/ System Updated Final Safety Analysis Report Rev. 9, 03/19 I Heat Generation in Homogenized Basket:
The design. decay heat loading for the AHSM is 24 kW The heat generation rate ("ii) for the homogenized basket region is calculated as:
ANUH-01.0150 A.4.11-8 All changes on this page are Amendment 4.
Updated Advanced NUHOMSS' System Updated Final Safety Analysis Report Proprietary Information on Pages A.4.11-9 through A.4.11-12 Withheld Pursuant to 10 CFR 2.390 ANUH-01.0150 A.4.11-9 All changes on this page are Amendment 4.
Rev. 9, 03/19 I
Updated Advanced NUHOMSS' System Updated Final Safety Analysis Report Component AHSMRoof AHSM Front Wall ANUH-01.0150 Table A.4.11.2-1 AHSM lnso/ation Averaged over 24 Averaged over 24
- hrs, hrs, Btul(hr-in2)
Btul(hr-ft2) 0.852 122.7 0.213 30.7 A.4.11-13 All changes on this page are Amendment 4.
Rev. 9, 03/19 *1 Averaged over 24 hrs, Wlm2 696.3 174.2
Updated Advanced NUHOM~ System Updated Final Safety Analysis Report Load Case Operation No.
Condition 1
Off-Normal 2 (1)
Accident Note:
Table A.4.11.2-2 Design Load Cases Description Off-Normal Hot, Steady-state Blocked Inlet and Outlet Vents for25hours Daily Average Ambient Temperature (
0F) 107 107 (1)
Initial temperatures are taken from steady-state results of LC #1.
ANUH-01.0150 A.4.11-14 All changes on this page are Amendment 4.
Rev. 9, 03/19 I lnsolation Yes Yes
Updated Advanced NUHOMffll' System Updated Final Safety Analysis Report ANUH-01.0150 Proprietary Information on This Page Withheld Pursuant to 10 CFR 2.390 A.4.11-15 All changes on this page are Amendment 4.
Rev. 9, 03/19 I
Updated Advanced NUHOMSW System Updated Final Safety Analysis Report Rev. 9, 03/19 I Table A.4.11.2-4 Maximum Component Temperatures for Off-Normal Condition Off-Normal Temperatures (LC#l)
Maximum Component Maximum Temp Temperatures.from (OF)
Table A.4.4-3 (OF)
DSC Shell 469(1)
DSC Support Rail 291 Inner Heat Shield 324 Upper Heat Shield 286 Upper Back Wall 232 Concrete 242 (1) Reported from Table A.4. 1-2 ANUH-01.0150 A.4.11-16 All changes on this page are Amendment 4.
Updated Advanced NUHOMSw System Updated Final Safety Analysis Report Rev. 9, 03/19 I TableA.4.11.2-5 Maximum Component Temperatures for Blocked Vent Accident Condition Maximum Maximum Maximum Temperatures Temperatures Temperatures Table Component (LC#2) at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (LC#2) at 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> A.4.1-3 (OF)
(OF) at 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> (OF)
DSC Shell 642 DSC Support Rail 615(1)
Inner Heat Shield 542(1)
Upper Heat Shield 542(1)
Upper Back Wall Concrete 392(1)
Thermocouple Temperature (1)
Maximum temperatures are reported after 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of blocked vent accident condition.
ANUH-01.0150 A.4.11-17 All changes on this page are Amendment 4.
Updated Advanced NUHOMS'JJ System Updated Final Safety Analysis Report ANUH-01.0150 Proprietary Information on This Page Withheld Pursuant to 10 CFR 2.390 A.4.11-18 All changes on this page are Amendment 4.
Rev. 9, 03/19 I
Updated Advanced NUHOMS"' System Updated Final Safety Analysis Report Rev. 9, 03/19 I Figure A.4.11.2-2 Axial View of AHSM Mesh with 24PT4 DSC on the Symmetrical Mid-plane ANUH-01.0150 A.4.11-19 All changes on this page are Amendment 4.
Updated Advanced NUHOMs& System Updated Final Safety Analysis Report Rev. 9, 03/19 I Detailed view of mesh around Rail Figure A.4.11.2-3 Cross-Sectional View of the Half-Symmetrical AHSM Mesh with 24PT4 DSC ANUH-01.0150 A.4.11-20 All changes on this page are Amendment 4.
Updated Advanced NUHOMs" System Updated Final Safety Analysis Report Rev. 9, 03/19 I Component Temperatures 450 400 350 300
~
~ 250 ta..
CII ci. 200 E
~
150 100 osc_shell DSC Support rail inner heat shield 50 upper_heat_shield upper back wall 0
0 2000 4000 6000 8000 10000 Iterations Figure A.4.11.2-4 Temperature Monitor of Various Components for Steady-State Off-Normal Simulation ANUH-01.01 50 A.4. 11-21 All changes on this page are Amendment 4.
Updated Advanced NUHOMSiy System Updated Final Safety Analysis Report 6~ft:.iature 395 (F(
378 360 343 325 308 290 273 255 238 220 203 186 Temperature
,alls (Fl
~~
269 26 255 247 240 233 225 218 211 203 196 189 182 174 167 160 152 145 (a) 24PT4 DSC Shell (c) Transition Rail Temperature Inner and Upper heal llhielda (FJ lif Ill 2~
244 236 229 221 213 205 197 18Q 182 174 166 158 Inner Heat Shield Upper Heat Shield (b) Inner and Upper Heat Shield Layature 260 (F)
~~
236 m 213 205 197 189 181 173 166 ms 142 134 126 119 11 1 Figure A.4.11.2-5 (d) AHSM Rev. 9, 03/19 I Temperature Contour for AHSM Loaded with 24PT4 DSC at Steady-State Off-Normal Conditions ANUH-Ol.0150 A.4.11-22 All changes on this page are Amendment 4.
Updated Advanced NUHOMS& System Updated Final Safety Analysis Report 230 225 220 215
~ 210 QI...
- , i! 205 QI Cl.
E 200
{!!.
195 190 185 180 0
500 450 I 400
~ 350 I 0 ! 300
.JO
- , i! 250 QI ci. 200 E
{!!. 150 100 f--
so 1-0 0
Thermocouple Tern erature 5
10 15 20 Time (hrs)
Figure A.4.11.2-6 Temperature Rise of As-Built Thermocouple Component Temperatures DSC Shell temp Inner heat shield upper_heat_shield 5
10 DSC Support rail Upper back wall 15 Time (hrs) 20 Figure A.4.11.2-7 Rev. 9, 03/19 I 25 25 Temperature Monitor of Various Components for Off-Normal Accident Simulation AN UH-01.01 50 A.4.11-23 All changes on this page are Amendment 4.
Updated Advanced NUHOMSw System Updated Final Safety Analysis Report ANUH-01.0150 Proprietary Information on This Page Withheld Pursuant to 10 CFR 2.390 A.4.11-24 All changes on this page are Amendment 4.
Rev. 9, 03/19 I
Updated Advanced NUHOMS System Updated Final Safety Analysis Report Table A.5.1-2 Summary of AHSM Dose Rates Dose Rate Dose Rate (mrem/hr)
Rev. 9, 03/19
. Surface Component Maximum Average Gamma 143.881 +/- 15.5%(a)
Rear End of the Neutron 0.207 +/- 4.3%
N/A TSBA<bl Total 144.088 +/-
15.5%
Gamma 1.115 +/- 4.2%
0.085 +/- 3.3%
Back of the Rear Neutron 0.008 +/- 1.4%
l.05E-3 +/- 1.6%
Shield Walfb)
Total 1.123 +/- 4.1%
0.086 +/- 3.3%
Gamma 44.318 +/- 5.3%
2.154 +/- 2.9%
Front<
0l Neutron 0.838 +/- 1.1%
0.138 +/- 7.7%
Total 45.156 +/- 5.2%
2.292 +/- 2.8%
Gamma 149.298 +/- 4.5%
0.011 +/- 2.7%
Roofd)
Neutron 0.279 +/- 1.6%
0.001 +/- 7.5%
Total 149.577 +/- 4.5%
0.012 +/- 2.6%
Gamma 6.657 +/- 6.8%
0.474 +/- 5.6%
AHSMTopCe)
Neutron 0.016 +/- 1.8%
1.56E-3 +/- 1.5%
Total 6.673 +/- 6.8%
0.476 +/- 5.6%
Gamma 1.790 +/- 3.3%
0.309 +/- 2.0%
Side Neutron 0.074 +/- 3.6%
l.06E-3 +/- 1.4%
Total 1.865 +/- 3.1%
0.319 +/- 1.9%
(a)
Statistical one standard deviation uncertainty in the Monte Carlo calculation.
(b)
The maximum gamma dose rates on the rear concrete surface (of"top" model) but below the roof elevation are less than 0.2 mrem/hr and the maximum gamma dose rates on this surface above the rooflevel are about 1.12 mrem/hr; i.e., the dose rate above the roof drops off very rapidly with distance in x from the vent (note the dose rate near the edge of the vent is 144.1 mrem/hr).
(c) These maximum dose rates on the front of the AHSM are based on the results calculated just in front of the entrance of the bottom air inlet. The maximum dose rate around the door is 4.453 mrem/hr (gamma dose rate 3.909 mremlhr and neutron dose rate 0.544 mremlhr) from the "top" MCNP model as shown in Figure A. 5. 4-3.
(d)
The dose rates are calculated on top of the AHSM roof. The maximum dose rates on the roof are based on the dose rates just at the roof vent opening. Knowing dose rates just above the roof vent opening is important, since this area must be accessed to clean the vent screens, if debris accumulates on the screens. For dose rates in front of the Top Shield Block Assembly (TSBA), the "Roof' maximum dose rate is below 1.0 mrem/hr.
The average dose rates are calculated over the roof segment in front of the TSBA (before its -x side).
(e)
The dose rates are calculated on the plane enveloping the AHSM from the top. The average dose rate is calculated over the entire plane enveloping the AHSM from the top.
This dose rate is used for the site dose rate analysis. The location of the maximum dose rate is near the rear end of the TSBA (its +x side, the side facing rear of the AHSM).
Note: Gamma results include the dose rates from gammas produced from neutrons in the neutron calculation. These partial gamma dose rates and the neutron dose rates have been multiplied by [1/(1-k)=l/(1-0.45)=1.82] to conservatively include neutron multiplication from induced fissions in the source region containing damaged fuel rods.
Note: The averaged dose rates are calculated over the planes enveloping the AHSM geometry, while peak dose rates are for localized areas. The average dose rates are needed for the site dose rate analysis.
ANUH-01.0150 A.5.1-5 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I A.8.2 Procedures for Unloading the 24PT4-DSC The following section outlines the procedures for retrieving the 24PT4-DSC from the AHSM and for removing the fuel assemblies from the 24PT4-DSC. These procedures are provided as a guide and are not intended to be limiting if the licensee determines that alternate means are available to accomplish the same operational objective. A process flow diagram for the Advanced NUHOMS System retrieval is presented in Figure A.8.2-1.
A.8.2.1 24PT4-DSC Retrieval froin the AHSM No change to the 24PT1-DSC Retrieval from the AHSM section as described in Chapter 8, Section 8.2.1 of the UFSAR.
A.8.2.2 Removal of Fuel from the 24PT4-DSC When the 24PT4-DSC has been removed from the AHSM, there are several potential options for off-site shipment of the fuel. These options include, but are not limited to, shipping the 24PT4-DSC with fuel assemblies or removing the fuel from the 24PT4-DSC as described below.
It is preferred to ship the 24PT4-DSC intact to a reprocessing facility,' monitored retrievable storage facility or permanent geologic repository in a compatible shipping cask, such as the MP197, licensed under 10 CFR Part 71. However, there are several reasons why it may be necessary to remove fuel assemblies from the 24PT4-DSC during the time period when the spent fuel pool is available. These include off-site transport in a transport cask requiring an alternate canister configuration, return of fuel assemblies to a spent fuel pool, or placement of fuel assemblies in a different 24PT4-DSC. Other reasons might include removing fuel assemblies at the end of service life or for inspection following an accident as discussed in Chapter A.12.
If it becomes necessary to remove fuel from the 24PT4-DSC prior to off-site shipment, there are two basic options available at the ISFSI or reactor site. The fuel assemblies could be removed and reloaded into a shipping cask using dry transfer techniques, or if the applicant so desires, the initial fuel loading sequence could be reversed and the plant's spent fuel pool utilized, if available. Procedures for unloading the 24PT4-DSC in a fuel pool are presented here, however wet or dry unloading procedures are essentially identical to those of24PT4-DSC loading through the weld removal process (beginning of preparation to placement of the transfer cask in the fuel pool). Prior to opening the 24PT4-DSC, the following operations are to be performed.
- 1.
Transfer the transfer cask to the cask handling area inside the plant's fuel handling building.
- 2.
Position and ready the trailer for access by the crane.
- 3.
Attach the lifting yoke to the crane hook.
- 4.
Engage the lifting yoke with the trunnions of the transfer cask.
- 5.
Visually inspect the yoke lifting hooks to insure that they are properly aligned and engaged onto the transfer cask trunnions.
ANUH-01.0150 A.8.2-1 All changes on this page are Amendment 4.
Updated Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I
- 6.
Lift the transfer cask approximately one inch off the trunnion supports. Visually inspect the yoke lifting hooks to insure that they are properly positioned on the trunnions.*
- 7.
Move the crane in a horizontal motion while simultaneously raising the crane hook vertically and lift the transfer cask off the trailer. Move the transfer cask to the cask decontamination area.
- 8.
Lower the transfer cask into the cask decontamination area in the vertical position.
- 9.
Wash the transfer cask to remove any dirt which may have accumulated during the 24PT4-DSC unloading and transfer operations.
- 10.
Place scaffolding around the transfer cask so that any point on the surface of the transfer cask is accessible to handling personnel.
- 11.
Unbolt the transfer cask top cover plate.
- 12.
Connect the rigging cables to the transfer cask top cover plate and lift the cover plate from the transfer cask. Set the transfer cask cover plate aside and disconnect the lid lifting cables.
- 13.
Install temporary shielding to reduce personnel exposure as required. Fill the transfer cask/24PT4-DSC annulus with clean water and seal the annulus The process of unloading the 24PT4-DSC into the spent fuel pool, if available, is similar to that used for loading. Operations that involve opening the 24PT4-DSC described below are to be carefully controlled in accordance with plant procedures. These operations are to be performed under the site's standard health physics guidelines for welding, grinding, and handling of potentially highly contaminated equipment. These are to include the use of prudent housekeeping measures and monitoring of airborne particulates. Procedures may require personnel to perform the work using respirators or supplied air.
If fuel needs to be removed from the 24PT4-DSC, precautions must be taken for the presence of damaged or oxidized fuel and to prevent radiological exposure to personnel during this operation. If degraded fuel is suspected, additional measures appropriate for the specific conditions are to be planned, reviewed, and implemented to minimize exposures to workers and radiological releases to the environment. A sampling of the atmosphere within the 24PT4-DSC should be taken prior to inspection or removal of fuel.
If the work is performed outside the fuel handling building, a tent may be constructed over the work area which may be kept under a negative pressure to control airborne particulates. Any radioactive gas release will be Kr-85, which is not readily captured. Whether the krypton is vented through the plant stack or allowed to be released directly depends on the plant operating requirements.
ANUH-01.0150 A.8.2-2 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I Following opening of the 24PT4-DSC, it is to be filled with demineralized or pool water prior to placement in the spent fuel pool to prevent a sudden inrush of pool water. Parameters related to reflooding the 24PT4-DSC cavity are addressed in Chapter A.3. Place transfer cask into the pool. The fuel unloading procedures listed below will be governed by the plant operating license under 10 CFR Part 50, and assume the availability of the spent fuel pool. The generic procedures for these operations are as follows:
- 1.
Locate the siphon and vent port using the indications on the top cover plate. Place a portable drill press on top of the 24PT4-DSC. Align the drill over the siphon port.
- 2.
Place an exhaust hood or tent over the 24PT4-DSC, if necessary. The exhaust should be filtered or routed to the site radwaste system.
- 3.
Drill a hole through the top cover plate to expose the siphon port quick connect.
- 4.
Drill a second hole through the top cover plate to expose the vent port quick connect.
CAUTION: (a)' The water fill rate must be regulated during this refloodi'ng operation to ensure that the 24PT4-DSC vent pressure does not exceed 20 psig.
(b) Provide for continuous hydrogen monitoring of the 24PT4-DSC cavity atmosphere during all subsequent cutting operations to ensure that a safety limit of 2.4% hydrogen concentration is not exceeded. Purge with 2-3 psig helium (or any other inert medium) as necessary to maintain the hydrogen concentration safely below this limit.
- 5.
Obtain a sample of the 24PT4-DSC atmosphere (confirm acceptable hydrogen concentration). Fill the 24PT4-DSC with water from the fuel pool through the siphon port with the vent port open and routed to the plant's off-gas system.
- 6.
Place welding blankets around the transfer cask and scaffolding.
- 7.
Using plasma arc-gouging, a mechanical cutting system or other suitable means, remove the weld from the out.er top cover plate and 24PT4-DSC shell. A fire watch should be placed on the scaffolding with the welder, as appropriate. The exhaust system should be operating at all times.
- 8.
The material or waste from the cutting or grinding process should be treated and handled in accordance with the plant's low level waste procedures unless determined otherwise.
- 9.
Remove the top of the tent, if necessary.
- 10.
Remove the exhaust hood, if necessary.
- 11.
Remove the outer top cover plate.
ANUH-01.0150 A.8.2-3 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I A.12.3 Limiting Condition for Operation (LCO) and Surveillance Requirements (SR)
Applicability BASES LCOs LCO 3.0.1 LCO 3.0.2 ANUH-01.0150 LCO 3.0.1, 3.0.2 and 3.0.4 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.
LCO 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LCO is required to be met (i.e.,
when the DSC is in the specified conditions of the Applicability statement of each Specification).
LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered. The Required Actions establish those remedial measures that must be taken within specified Completion Times when
- the requirements of an LCO are not met. This Specification establishes that:
- a.
Completion of the Required Actions within the specified Completion Times constitutes compliance with a Specification; and
- b.
Completion of the Required Actions is not required when an LCO is met within the specified Completion Time, unless otherwise specified.
There are two basic types of Required Actions. The first type of Required Action specifies a time limit in which the LCO must be met. This time limit is the Completion Time to restore a system or component or to restore variables to within specifie~ limits. If this type of Required Action is not completed within the specified Completion Time, the DSC may have to be placed in the spent fuel pool, if available, and unloaded. (Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS.) The second type of Required Action specifies the remedial measures that permit continued operation of the unit that is not further restricted by the Completion Time. In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation.
Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications.
A.12-4 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I A.14 DECOMMISSIONING A.14.1 Decommissioning Considerations The Advanced NUHOMS System design features include inherent ease and simplicity for decommissioning by providing easily decontaminable surfaces and isolating the external surfaces of the 24PT4-DSC from contact with the fuel pool. At the end of its service life, the 24PT4-DSC decommissioning could be performed by one of the options listed below:
Option 1, the 24PT4-DSC, including stored spent fuel, could be shipped to either a monitored retrievable storage system (MRS) or a geological repository for final disposal, or Option 2, the spent fuel could be removed from the 24PT4-DSC (in the spent fuel pool, if still available onsite, or using dry transfer techniques or other means) and the fuel shipped offsite in an NRC approved transportation cask.
The first option requires that the Part 72 24PT4-DSC (i.e., designed for storage) be upgraded to current Part 71 regulations. An amendment to the MP197 CoC [A14.2] will be initiated to allow for transport of the 24PT4-DSC using the MP197 cask.
The first option does not require any decommissioning of the 24PT4-DSC. No residual contami-nation is expected to be left behind on the concrete AHSM. The AHSM, fence, and peripheral utility structures will require no decontamination or special handling after the last 24PT4-DSC is removed. The AHSM, fence, and peripheral utility structures could be demolished and recycled
- with normal construction techniques.
The second option, which assumes the availability of a spent fuel pool onsite, would require decontamination of the 24PT4-DSC and transfer cask (if applicable). The sources of contamination in the interior of the 24PT4-DSC or transfer cask would be the primary contamination left from the spent fuel pool water, if unloading using the spent fuel pool; or crud, hot particles and fines from the spent fuel pins. This contamination could be removed with a high pressure water spray. If further surface decontamination of the 24PT4-DSC or transfer cask is necessary, electro-polishing or chemical etching can be used to clean the contaminated surface. After decontamination, the 24PT4-DSC and/or transfer cask could be cut up for scrap, partially scrapped, or refurbished for reuse. Any activated metal would be shipped as low level radioactive waste to a disposal facility.
A review of cask activation analyses previously performed for similar systems (TN-32 cask
[A14.4] and NUHOMS site license storage system) indicates that the levels of activation of the 24PT4-DSC, AHSM and transfer cask would be orders of magnitude below the specific activity of the isotopes listed in Tab~es 1 and 2 of 10 CPR 61.55 [A14.3]. A detailed analysis is not considered necessary based on the significant margins determined from these analyses. A comparison of the source terms for this application to those referenced above including the activation analysis summary for the above applications is provided below:
ANUH-01.0150 A.14.1-1 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I Comparison of Source Terms for Activation Analyses Source Term 24PT4-DSC TN-32 (Metal Cask)
NUHOMS Site License HSM y (y/sec/assy) 7.501 X 1015 5.3 X 1015 1.53 X 1015 n (n/sec/assy) 3.696 X 108 3.3 ~ 108 2.23x108 TN 32 and NUHOMS Site License HSM Activation Analysis Results Activity Ci/m 3
Nuclide HSM HSM Steel TN-32 10 CFR 61.55 Concrete Limit H-3 8.3 X 10-11 40 C-14 2.3 X 10-10 8
Co-60 4.4 X 10-5 8.1 X 10-2 7.7 X 10-6 700 Ni-59 1.4x10-10 3.1 X 10-6 2.5 X 10-6 220 Ni-63 8.3 X 10-8 3.2 X 10-4 3.4 X 10-4 3.5 Nb-94 3.9 X 10-8 0.2
<5 year 4.6 X 10-3 2.0x10-1 2.3 X 10-2 700 half life Following surface decontamination, the radiation levels in the 24PT4-DSC or transfer cask due to activation will be below the acceptable limits ofRegulatory Guide 1.86 [A14.1]. The activation levels of the 24PT4-DSC or transfer cask materials will be far below the specific activity limits for both short and long lived nuclides for Class A waste. A detailed evaluation will be performed at the time of decommissioning to determine the appropriate mode of disposal, should refurbishment not be elected.
The procedure for decommissioning a 24PT4-DSC or transfer cask not being returned to service is summarized below:
Remove fuel in accordance with the unloading procedures of Chapter A.8.
Survey interior of 24PT4-DSC or transfer cask. If the spent fuel pool is available, wash down the inside of the 24PT4-DSC or transfer cask. Pump out and filter contaminated water and cleaning agent. Survey interior of24PT4-DSC or transfer cask again, decontaminate as required. It is expected that surface contamination will be minimal. If so, dispose of the 24PT4-DSC or transfer cask body as scrap metal. If unable to decontaminate to acceptable levels, the 24PT4-DSC and/or transfer cask body can be disposed of as low level radioactive waste.
ANUH-01.0150 A.14.1-2 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I B.4.8 Thermal Evaluation for Loading/Unloading Conditions All individual fuel assembly loading operations occur when the 32PTH2 DSC and OS200FC TC are in the spent fuel pool. The fuel is always submerged in free-flowing pool water permitting heat dissipation. After completion of the fuel loading, the TC and DSC are removed from the pool and the DSC is drained, dried, sealed, and backfilled with helium. These operations occur when the annulus between the TC and DSC remains filled with water.
The water in the annulus is monitored and replenished with fresh water to prevent boiling and maintain the water level if excessive evaporation occurs as noted for the fuel loading operation procedures in Sections B.8.1.1.3 and B.8.1.1.4. Presence of water within the annulus maintains the maximum DSC shell temperature below the boiling temperature of water in open atmosphere (212 °F).
Water in the DSC cavity is forced out of the cavity (blowdown operation) before the start of vacuum drying. Helium is used as the medium to remove water and subsequent vacuum drying occurs with a helium environment in the DSC cavity. The vacuum drying operation does not reduce the pressure sufficiently to reduce the thermal conductivity of.the helium in the DSC cavity as discussed in Appendix U, Section U.4.7.1 of the UFSAR for the Standardized NUHOMS System [B4.22].
With helium being present during vacuum drying operations and a DSC shell temperature equal to water boiling temperature of2I2 °F, the 32PTH2 DSC model described in Section B.4.6.2.1 is used in a steady-state analysis to determine the maximum fuel cladding temperature for vacuum drying operations. The maximum fuel cladding temperature for vacuum drying operations in the 32PTH2 DSC is 572 °F and 540 °F for 37.2 kW and 32.0 kW decay heat loads, respectively.
The presence of helium during blowdown and vacuum drying operations eliminates the thermal cycling of fuel cladding during helium backfilling of the DSCs subsequent to vacuum drying.
Therefore, the thermal cycling limit of 65 °C (117 °F) for short-term operations set by NUREG-1536 [B4.3] is satisfied for vacuum drying operation.
The bounding unloading operation considered is the reflood of the 32PTH2 DSCs with water during the time period when the spent fuel pool is available. For unloading operations, the DSC is filled with the spent fuel pool water through its siphon port. During this filling operation, the 32PTH2 DSC vent port remains open with effluents routed to the plant's off-gas monitoring system.
The maximum fuel cladding temperature during the reflooding event is significantly less than the.
vacuum drying condition owing to the presence of water/steam in the DSC cavity. Based on the above rationale, the maximum cladding temperature during unloading operation is bounded by the maximum fuel cladding temperature for vacuum drying operation.
ANUH-01.0150 B.4.8-1 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I B.8.2 Procedures for Unloading the 32PTH2 DSC The following section outlines the procedures for retrieving the 32PTH2 DSC from the AHSM-HS and for removing the fuel assemblies from the 32PTH2 DSC. These procedures are provided as a guide and are not intended to be limiting if the licensee determines that alternate means are available to accomplish the same operational objective. A flow chart of the unloading operations of the 32PTH2 system is provided in Figure B.8.2-1.
B.8.2.1 32PTH2 DSC Retrieval from the AHSM-HS No change to the 24PT1-DSC retrieval from the AHSM section as described in Chapter 8, Section 8.2.1 or the 24PT4-DSC retrieval described in Appendix A, Section A.8.2.1.
The retrieval of the 32PTH2 DSC from the AHSM-HS is, however, outlined in Figure B.8.2-1 below.
B.8.2.2 Removal of Fuel from the 32PTH2 DSC When the 32PTH2 DSC has been removed from the AHSM-HS, there are several potential options for off-site shipment of the fuel. These options include, but are not limited to, shipping the 32PTH2 DSC with fuel assemblies or removing the fuel from the 32PTH2 DSC as described below. It is preferred to ship the 32PTH2 DSC intact to a reprocessing facility, monitored retrievable storage facility or permanent geologic repository in a compatible transportation packaging licensed under 10 CPR Part 71.
If it becomes necessary to remove fuel from the 32PTH2 DSC prior to off-site shipment, there are two basic options available at the ISFSI or reactor site. The fuel assemblies could be removed and reloaded into a licensed transport packaging using dry transfer techniques, or if the applicant so desires, the initial fuel loading sequence could be reversed and the plant's spent fuel pool utilized, if available. Procedures for unloading the 32PTH2 DSC in a fuel pool are presented here, however wet or dry unloading procedures are essentially identical to those of 32PTH2 DSC loading through the weld removal process (beginning of preparation to placement of the TC in the fuel pool). Prior to opening the 32PTH2 DSC, the following operations are to be performed.
- 1.
Transfer the TC to the TC handling area inside the plant's fuel handling building.
- 2.
Position and ready the trailer for access by the crane.
- 3.
Attach the lifting yoke to the crane hook.
- 4.
Engage the lifting yoke with the trunnions of the TC.
- 5.
Visually inspect the yoke lifting hooks to insure that they are properly aligned and engaged onto the TC trunnions.
- 6.
Lift the TC approximately one inch off the upper trunnion supports.
- 7.
Translate the crane horizontally while simultaneously raising the crane hook vertically and lift the TC off the trailer. Move the TC to the TC decontamination area.
- 8.
Lower the TC into the TC decontamination area in the vertical position.
ANUH-01.0150 B.8.2-1 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19
- 9.
Wash the TC to remove any dirt which may have accumulated during the 32PTH2 DSC unloading and transfer operations.
- 10.
Place scaffolding around the TC so that any point on the surface of the TC is accessible to handling personnel.
- 11.
Unbolt the TC cover plate assembly.
- 12.
Connect the rigging cables to the TC cover plate assembly and lift it from the TC. Set the TC cover plate assembly aside and disconnect the lid lifting cables.
- 13.
Install temporary shielding to reduce personnel exposure as required. Fill the TC/32PTH2 DSC annulus with clean water and install a protective cover for the annulus.
The process of unloading the 32PTH2 DSC into the spent fuel pool is similar to that used for loading. Operations that involve opening the 32PTH2 DSC described below are to be carefully controlled in accordance with plant procedures. These operations are to be performed under the site's standard health physics guidelines for welding, grinding, and handling of potentially highly contaminated equipment. These are to include the use of prudent housekeeping measures and monitoring of ai.rbome particulates. Procedures may require personnel to perform the work using respirators or supplied air.
If fuel needs to be removed from the 32PTH2 DSC, precautions must be taken for the presence of damaged or oxidized fuel and to prevent radiological exposure to personnel during this operation. If degraded fuel is suspected, additional measures appropriate for the specific conditions are to be planned, reviewed, and implemented to minimize exposures to workers and radiological releases to the environment. A sampling of the atmosphere within the 32PTH2 DSC is required prior to inspection or removal of fuel per Technical Specification 5.1.
If the work is performed outside the fuel handling building, a tent may be constructed over the work area which may be kept under a negative pressure to control airborne particulates. Any radioactive gas release will be Kr-85, which is not readily captured. Whether the krypton is vented through the plant stack or allowed to be released directly depends on the plant operating requirements.
Following opening of the 32PTH2 DSC, it is filled with demineralized water or pool water prior to placement in the spent fuel pool to prevent a sudden inrush of pool water. Parameters related to reflooding the 32PTH2 DSC cavity are addressed in Chapter 3. Place transfer cask into the pool. The fuel unloading procedures listed below will be governed by the plant operating license under 10 CFR Part 50, if this license is still active. The generic procedures for these operations are as follows:
- 1.
Locate the siphon and vent ports using the indications on the outer top cover plate. Place a portable drill press on top of the 32PTH2 DSC. Align the drill over the siphon port.
- 2.
Place an exhaust hood or tent over the 32PTH2 DSC, if necessary. The exhaust should be filtered or routed to the site radwaste system.
CAUTION: Radiation dose rates are expected to be high at the vent and siphon port location. Use proper ALARA practices ( e.g., use of temporary shielding, appropriate positioning of personnel, etc.) to minimize personnel exposure.
ANUH-01.0150 B.8.2-2 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I
- 3.
Drill holes through the outer top cover plate and siphon port cover plate to expose the siphon port quick connect.
- 4.
Drill holes through the outer top cover plate and vent port cover plate to expose the vent port quick connect.
ANUH-01.0150 B.8.2-2a All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19
- 5.
Obtain a sample of the 32PTH2 DSC _atmosphere per the requirements of Technical Specification 5.1 (confirm acceptable hydrogen concentration). Fill the 32PTH2 DSC with water from the fuel pool through the siphon port with the vent port open and routed to the plant's off-gas system.
CAUTION:
- a. The water fill rate must be regulated during this reflooding operation to ensure that the 32PTH2 DSC vent pressure does not exceed 20 psig.
- b. Per Technical Specification 5.2.6, provide for continuous hydrogen monitoring of the 32PTH2 DSC cavity atmosphere during all subsequent cutting operations to ensure that a safety limit of 2.4% hydrogen concentration is not exceeded [B8.2] and [B8.3].
Purge with 1-3 psig helium as necessary to maintain the hydrogen concentration safely below this limit.
- 6.
Using plasma arc-gouging, a mechanical cutting system or other suitable means, remove the weld from the outer top cover plate and the 32PTH2 DSC shell and remove the outer top cover plate. The exhaust system should be operating at all times.
- 7.
The material or waste from the cutting or grinding process should be treated and handled in accordance with the plant's low level waste procedures unless determined otherwise.
- 8.
Using plasma arc-gouging, a mechanical cutting system or other suitable means, remove the weld from the inner top cover plate and the 32PTH2 DSC shell in the same manner as the outer top cover plate. Remove the inner top cover plate. Remove any remaining excess material on the inside shell surface by grinding.
- 9.
Clean the TC surface of dirt and any debris which may be on the TC surface as a result of the weld removal operation. Any other procedures which are required for the operation of the TC, including installation of the annulus seal, should take place at this point as necessary.
- 10.
Engage the trunnions with the yoke, install eyebolts into the top shield plug and connect the rigging cables to the eyebolts.
- 11.
Visually inspect the lifting hooks of the yoke to insure that they are properly positioned on the trunnions.
- 12.
The TC should be lifted just high enough to allow the weight of the TC to be distributed onto the yoke lifting hooks. Inspect the lifting hooks to insure that they are properly positioned on the trunnions.
- 13.
Install suitable protective material onto the bottom of the TC to minimize TC contamination, as appropriate. Move the TC to the spent fuel pool.
- 14.
Prior to lowering the TC into the pool, adjust the pool water level, if necessary, to accommodate the volume of water which will be displaced by the TC during the operation.
- 15.
Position the TC over the designated area in the fuel pool
- 16.
Lower the TC into the pool. As the TC is being lowered, the exterior surface of the TC should be sprayed with clean demineralized water.
ANUH-01.0150 B.8.2-3 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19
- 17.
Lower the TC into the fuel pool leaving the top surface of the TC approximately one foot above the surface of the pool water. Verify correct connections of the annulus seal and annulus/neutron shield tanks if used.
- 18.
Fill the top of the 32PTH2 DSC with water as needed and continue lowering the TC into the pool.
- 19.
Disengage the lifting yoke from the TC and lift the top shield plug from the 32PTH2 DSC.
- 20.
If the 32PTH2 DSC contains damaged fuel assemblies, remove the top end caps.
Remove the fuel from the 32PTH2 DSC and place the fuel into the spent fuel racks.
- 21.
Lower the top shield plug into the empty 32PTH2 DSC (optional).
- 22.
Visually verify that the top shield plug is properly positioned, if necessary.
- 23.
Engage the lifting yoke onto the TC trunnions.
- 24.
Visually verify that the yoke lifting hooks are properly engaged with the TC trunnions.
- 25.
Lift the TC by a small amount and verify that the lifting hooks are properly engaged with the trunnions.
- 26.
Lift the TC to the pool surface. Prior to raising the top of the TC above the water surface, stop vertical movement and inspect the top shield plug to ensure that it is properly positioned. If the top shield plug is not properly seated, lower the TC back to the fuel pool and reposition the plug.
- 27.
As the TC is raised above the pool surface, drain the excess water from the 32PTH2 DSC above the top shield plug back into the fuel pool.
- 28.
Lift the TC from the pool. As the TC is rising out of the pool, spray the exposed portion of the TC with demineralized water.
- 29.
Move the TC to the TC decontamination area.
- 30.
Check radiation levels around the perimeter of the TC. The TC exterior surface should be decontaminated, if necessary.
- 31.
Place scaffolding around the TC so that any point on the surface of the TC is easily accessible to personnel.
- 32.
Connect a water draining/pumping device to the siphon port of the 32PTH2 DSC and remove water from the 32PTH2 DSC cavity.
- 33.
The top cover plates may be welded into place as required.
- 34.
Decontaminate the 32PTH2 DSC, as necessary, and handle in accordance with low-level waste procedures. Alternatively, the 32PTH2 DSC may be repaired and recertified for reuse.
ANUH-01.0150 B.8.2-4 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I B.12.3 Limiting Condition for Operation (LCO) and Surveillance Requirement (SR)
Applicability BASES LCOs LCO 3.0.1 LCO 3.0.2 ANUH-01.0150 LCO 3.0.1, 3.0.2, and 3.0.4 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.
LCO 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LCO is required to be met (i.e.,
when the DSC is in the specified conditions of the Applicability statement of each Specification).
LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered. The Required Actions establish those remedial measures that must be taken within specified Completion Times when the requirements of an LCO are not met. This Specification establishes that:
- a. Completion of the Required Actions within the specified Completion Times constitutes compliance with a Specification; and
- b. Completion of the Required Actions is not required when an LCO is met within the specified Completion Time, unless otherwise specified.
There are two basic types of Required Actions. The first type of Required Action specifies a time limit in which the LCO must be met. This time limit is the Completion Time to restore a system or component or to restore variables to within specified limits. If this type of Required Action is not completed within the specified Completion Time, the DSC may have to be placed in the spent fuel pool, if available, and unloaded. (Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS.) The second type of Required Action specifies the remedial measures that permit continued operation of the unit that is not further restricted by the Completion Time. In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation.
Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications.
B.12-5 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I B.14 DECOMMISSIONING B.14.1 Decommissioning Considerations The NUHOMS 32PTH2 system design features include inherent ease and simplicity for decommissioning by providing easily decontaminated surfaces and isolating the external surfaces of the 32PTH2 DSC from contact with the fuel pool. At the end of its service life, the 32PTH2 DSC decommissioning could be performed by one of the options listed below:
Option 1, the 32PTH2 DSC, including stored spent fuel, could be shipped to either a monitored retrievable storage system (MRS) or a geological repository for final disposal, or Option 2, the spent fuel could be removed from the 32PTH2 DSC in the spent fuel pool, if still available onsite, or using dry transfer techniques or other means) and shipped in an NRC approved transportation cask.
The first option requires that the 32PTH2 DSC designed for storage under 10 CFR Part 72, be upgraded to current 10 CFR Part 71 regulations. An amendment to the MP197 CoC [B14.2] will be initiated to allow for transport of the 32PTH2 DSC using the MP197HB cask.
The first option does not require any decommissioning of the 32PTH2 DSC. No residual contamination is expected to be left behind on the concrete AHSM-HS. The AHSM-HS, fence, and peripheral utility structures will require no decontamination or special handling after the last 32PTH2 DSC is removed. The AHSM-HS, fence, and peripheral utility structures could be demolished and recycled with normal construction techniques.
The second option, which assumes the availability of a spent fuel pool onsite, would require decontamination of the 32PTH2 DSC and transfer cask (if applicable). The sources of contamination in the interior of the 32PTH2 DSC or transfer cask would be the primary contamination left from the spent fuel pool water if unloading using the spent fuel pool.
Additionally, there could be crud, hot particles and fines from the spent fuel rods. This contamination could be removed with a high pressure water spray. If further surface decontamination of the 32PTH2 DSC or transfer cask is necessary, electro-polishing or chemical etching can be used to clean the contaminated surface. After decontamination, the 32PTH2 DSC and/or transfer cask could be cut up for scrap, partially scrapped, or refurbished for reuse. Any activated metal would be shipped as low level radioactive waste to a disposal facility.
A review of cask activation analyses previously performed for similar systems (TN-32 cask
[B 14.4] and NUHOMS site license storage system) indicates that the levels of activation of the 32PTH2 DSC, AHSM-HS and transfer cask would be orders of magnitude below the specific activity of the isotopes listed in Tables 1 and 2 of 10 CFR 61.55 [B14.3]. A detailed analysis is not considered necessary based on the significant margins determined from these analyses. A comparison of the source terms for this application to those referenced above, including the activation analysis summary for the above applications, is provided below:
ANUH-01.0150 B.14.1-1 All changes on this page are Amendment 4.
Advanced NUHOMS System Updated Final Safety Analysis Report Rev. 9, 03/19 I Comparison of Source Terms for Activation Analyses Source Term 32PTH2 DSC TN-32 (Metal Cask)
NUHOMS Site License HSM y (y/sec/assy) 6.3 X 1015 5.3 X 1015 1.53x1015
- n (n/sec/assy) 4.2 X 108 3.3 X 108 2.23 X 108 TN-32 and NUHOMS Site License HSM Activation Analysis Results Activity Ci/m 3
Nuclide HSM 10 CFR 61.55 Concrete HSM Steel TN-32 Limit H-3 8.3 X 10-11 40 C-14 2.3 X 10-10 8
Co-60 4.4 X 10-5 8.1 X 10-2 7.7 X 10-6 700 Ni-59 1.4 X 10-1o 3.1 X 10-6 2.5 X 10-6 220 Ni-63 8.3 X 10-8 3.2 X 10-4 3.4 X 10-4 3.5 Nb-94 3.9 X 10-8 0.2
<5 year half life 4.6 X 10-3 2.0 X 10-1 2.3 X 10-2 700 Following surface decontamination, the radiation levels in the 32PTH2 DSC or transfer cask due to activation will be below the acceptable limits of Regulatory Guide 1.86 [B14.1]. The activation levels of the 32PTH2 DSC or transfer cask materials will be far below the specific activity limits for both short and long lived nuclides fo,r Class A waste. A detailed evaluation will be performed at the time of decommissioning to determine the appropriate mode of disposal, should refurbishment not be elected.
The procedure for decommissioning a 32PTH2 DSC or transfer cask not being returned to service is summarized below:
Remove fuel in accordance with the unloading procedures of Chapter B.8.
Survey interior of 32PTH2 DSC or transfer cask. If the spent fuel pool is available, wash down the inside of the 32PTH2 DSC or transfer cask. Pump out and filter contaminated water and cleaning agent. Survey interior of 32PTH2 DSC or transfer cask again, decontaminate as required. It is expected that surface contamination will be minimal. If so, dispose of the 32PTH2 DSC or transfer cask body as scrap metal. If unable to decontaminate to acceptable levels, the 32PTH2 DSC and/or transfer cask body can be disposed of as low level radioactive waste.
Decontaminate the top shield plug assembly and top cover plates until able to dispose of as scrap metal. If unable to achieve acceptable levels, dispose of these items as low level radioactive waste.
ANUH-01.0150 B.14.1-2 All changes on this page are Amendment 4.