ML24215A078
| ML24215A078 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 08/02/2024 |
| From: | NuScale |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24215A000 | List:
|
| References | |
| LO-169995 | |
| Download: ML24215A078 (1) | |
Text
Response to SDAA Audit Question Question Number: A-3.11-6 Receipt Date: 01/02/2024 Question:
COL Item 3.11-2 states that, An applicant that references the NuScale Power Plant US460 standard design will ensure the Environmental Qualification Program cited in COL Item 3.11-1 includes a description of how equipment located in harsh conditions will be monitored and managed throughout plant life. This description will include methodology to ensure equipment located in harsh environments will remain qualified if the measured dose is higher than the calculated dose.
The description for monitoring equipment located in harsh conditions described in COL Item 3.11-2 only requires the COL applicant to consider radiation dose. However, there is no other COL items related to monitoring equipment for other environmental parameters described in 10 CFR 50.49(e), such as temperature, pressure, humidity and chemical expose. Explain how other environmental parameters described in CFR 50.49(e) will be monitored throughout the life of the equipment.
Response
Combined license (COL) Item 3.11-2 is revised to include the environmental parameters temperature, pressure, humidity, radiation and chemical exposure. The revised COL Item 3.11-2 wording is as follows:
An applicant that references the NuScale Power Plant US460 standard design will ensure the Environmental Qualification Program cited in COL Item 3.11-1 includes a description of how equipment subject to program requirements will be monitored and managed throughout plant life. This description will include methodology to ensure equipment located in harsh or mild environments will remain qualified if an actual NuScale Nonproprietary NuScale Nonproprietary
environmental parameter, such as temperature, pressure, humidity, radiation, or chemical exposure deviates from the acceptable range for which the component is qualified.
The revised COL Item 3.11-2 wording also contains revisions based on the response to Request for Additional Information question 12.3-1.
Markups of the affected changes, as described in the response, are provided below:
NuScale Nonproprietary NuScale Nonproprietary
NuScale Final Safety Analysis Report Interfaces with Standard Design NuScale US460 SDAA 1.8-6 Draft Revision 2 COL Item 3.9-4:
An applicant that references the NuScale Power Plant US460 standard design will provide applicable test procedures before the start of testing and will submit test and inspection results from the Comprehensive Vibration Assessment Program for the NuScale Power Module in accordance with Regulatory Guide 1.20.
3.9 COL Item 3.9-5:
An applicant that references the NuScale Power Plant US460 standard design will implement a control rod drive system Operability Assurance Program that meets the requirements described in Section 3.9.4.4 and provide a summary of the testing program and results.
3.9 COL Item 3.9-6:
An applicant that references the NuScale Power Plant US460 standard design will develop a Reactor Vessel Internals Reliability Program to address industry identified aging degradation mechanism issues.
3.9 COL Item 3.9-7:
An applicant that references the NuScale Power Plant US460 standard design will provide a summary of reactor core support structure American Society of Mechanical Engineers (ASME) service level stresses, deformation, and cumulative usage factor values for each component and each operating condition in conformance with ASME Boiler and Pressure Vessel Code Section III Subsection NG.
3.9 COL Item 3.9-8:
An applicant that references the NuScale Power Plant US460 standard design will establish Preservice and Inservice Testing Programs. These programs are to be consistent with the requirements in the latest edition and addenda of the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code incorporated by reference in 10 CFR 50.55a.
3.9 COL Item 3.9-9:
An applicant that references the NuScale Power Plant US460 standard design will develop specific test procedures to allow detection and monitoring of power-operated valve assembly performance sufficient to satisfy periodic verification design basis capability requirements.
3.9 COL Item 3.9-10:
An applicant that references the NuScale Power Plant US460 standard design will develop specific test procedures to allow detection and monitoring of emergency core cooling system valve assembly performance sufficient to satisfy periodic verification of design-basis capability requirements.
3.9 COL Item 3.11-1:
An applicant that references the NuScale Power Plant US460 standard design will submit a full description of the Environmental Qualification Program and milestones and completion dates for program implementation.
3.11 RAI 12.3-1 COL Item 3.11-2:
An applicant that references the NuScale Power Plant US460 standard design will ensure the Environmental Qualification Program cited in COL Item 3.11-1 includes a description of how equipment located in harsh conditionssubject to program requirements will be monitored and managed throughout plant life. This description will include methodology to ensure equipment located in harsh or mild environments will remain qualified if an actual environment parameter, such as temperature, pressure, humidity, radiation, or chemical exposure deviates from the acceptable range for which the component is qualified.the measured dose is higher than the calculated dose.
3.11 COL Item 3.11-3:
An applicant that references the NuScale Power Plant US460 standard design will implement an Environmental Qualification Operational Program that incorporates the aspects in Section 3.11.5 specific to the environmental qualification of mechanical and electrical equipment. This program will include an update to Table 3.11-1 to include commodities that support equipment listed in Table 3.11-1.
3.11 COL Item 3.12-1:
An applicant that references the NuScale Power Plant US460 standard design may use a piping analysis program other than the programs listed in Section 3.12.4; however, the applicant will implement a benchmark program using the models for the NuScale Power Plant US460 standard design.
3.12 COL Item 3.12-2:
An applicant that references the NuScale Power Plant US460 standard design will confirm that the site-specific seismic response is within the parameters specified in Section 3.7. An applicant may perform a site-specific piping stress analysis in accordance with the methodologies described in this section, as appropriate.
3.12 Table 1.8-1: Combined License Information Items (Continued)
Item No.
Description of COL Information Item Section
NuScale Final Safety Analysis Report Environmental Qualification of Mechanical and Electrical Equipment NuScale US460 SDAA 3.11-8 Draft Revision 2 equipment listed in Table 3.11-1 is evaluated to ensure it remains qualified.
Section 12.3 discusses normal operational dose rates.
The normal operations dose rates for environmental qualification are derived from direct gamma radiation emitted by radioactive fluids. Beta radiation and Bremsstrahlung radiation during normal operations are considered negligible contributors to doses in comparison to the gamma radiation and therefore are omitted. Normal doses within the CNV and other areas also account for neutron fluence, when applicable, by equating the neutron fluence to an equivalent dose in rads.
Accident dose rates include a submersion dose and a direct dose contribution.
The submersion dose is derived from both the gamma and beta radiation. Beta radiation is attenuated by low-density equipment enclosures. Alpha radiation is neglected from both the normal and accident environmental qualification dose rates because the alpha particle is easily attenuated by air.
Reference 3.11-10 and Section 12.2.1 present the methodology that forms the basis for the accident dose rate calculations. Section 15.0.3 discusses the assumptions associated with the accident dose rates. Appendix 3C provides additional information on normal and accident dose rates used for environmental qualification.
RAI 12.3-1 COL Item 3.11-2: An applicant that references the NuScale Power Plant US460 standard design will ensure the Environmental Qualification Program cited in COL Item 3.11-1 includes a description of how equipment subject to program requirementslocated in harsh conditions will be monitored and managed throughout plant life. This description will include methodology to ensure equipment located in harsh or mild environments will remain qualified if an actual environment parameter, such as temperature, pressure, humidity, radiation, or chemical exposure deviates from the acceptable range for which the component is qualified.the measured dose is higher than the calculated dose.
3.11.5 Environmental Qualification Operational Program An Environmental Qualification Operational Program ensures continued capability of qualified mechanical and electrical equipment to perform its design function throughout its qualified life. The Environmental Qualification Operational Program contains the following aspects specific to the environmental qualification of mechanical and electrical equipment:
evaluation of environmental qualification results to establish activities to support continued environmental qualification for the entire time an item is installed in the
- plant, determination of surveillance and preventive maintenance activities based on environmental qualification results,
NuScale Final Safety Analysis Report Radiation Protection Design Features NuScale US460 SDAA 12.3-9 Draft Revision 2 Dose is limited to protect plant personnel, members of the public, and susceptible equipment subject to environmental qualification requirements.
Shielding performance is in accordance with the following criteria:
ALARA radiation protection principles of 10 CFR 20 exposure limits of 10 CFR 20 dose limits of principal design criteria (PDC) 19 In addition, plant layout and shielding are used to limit equipment radiation doses to levels that are consistent with the assumptions used to demonstrate environmental qualification.
12.3.2.2 Design Considerations Shielding is provided for radioactive systems and components to reduce radiation levels commensurate with area personnel access requirements and ALARA principles. Section 12.3.1 describes the radiation zones and indicates the radiation levels for those plant areas.
Section 12.3.1 describes shielding design features including permanent shielding and separation of components that constitute substantial radiation sources, the use of shielded cubicles, labyrinths, and shielded entrances to minimize radiation exposures. The selection of shielding materials considers the ambient environment and potential degradation mechanisms. Temporary shielding is considered where it is impractical to provide permanent shielding for substantial radiation sources.
Consistent with RG 8.8, streaming of radiation into accessible areas through penetrations for pipes, ducts, and other shield discontinuities is reduced by using layouts that prevent alignment with the radiation source, placing penetrations above head height to reduce personnel exposures, and using shadow shields to attenuate radiation streaming.
Consistent with RG 8.8, shielding analysis employs accurate modeling techniques and conservative approaches in the determination of shielding thickness. Source terms, geometries, and field intensities are analyzed conservatively. In addition to normal conditions, source terms include transient conditions such as resin transfers.
RAI 12.3-1 The material used for a significant portion of plant shielding is concrete. For most applications, concrete shielding is designed in accordance with American National Standards Institute/American Nuclear Society (ANSI/ANS) 6.4-2006 (Reference 12.3-1). Table 12.3-5 and Table 12.3-6 show the nominal concrete equivalentshielding thicknesses gamma attenuation assumed in the shielding analyses in plant buildings. In addition to concrete, other types of materials such as steel, water, tungsten, and polymer composites are considered for both permanent and temporary shielding. The use of lead is minimized.
NuScale Final Safety Analysis Report Radiation Protection Design Features NuScale US460 SDAA 12.3-10 Draft Revision 2 A listing of radiation shield barrier equivalent doors is provided in Table 12.3-7 for the RWB. There are no credited shield doors for the RXB. They are modeled as open doorways.
Shield floor plugs provide an equivalent radiation attenuation as the shield floor that contains the plug.
12.3.2.3 Calculation Methods The primary computer program to evaluate shielding is Monte Carlo N-Particle Transport Code (MCNP6) (Reference 12.3-2), which was developed by Los Alamos National Laboratory. The MCNP6 code is used for shielding calculations and for dose rate determinations.
Radioactive components in the RXB and RWB are modeled using MCNP6.
Section 12.2 describes the codes used to prepare source strength input data. A three-dimensional shielding model is constructed for radioactive components using structure, location, and equipment data. Source geometries and source term distributions and intensities are conservatively determined. In general, the component source geometries are modeled as cylindrical volumes that incorporate the full volume of the component.
RAI 12.3-1 Shielding credit and material selections for MCNP6 cells are conservatively applied. Credit is not taken for reinforcing steel bars in the concrete. Table 12.3-5 and Table 12.3-6 describe the credited shield barriers for the RWB and RXB in terms of nominal concrete attenuationequivalent thicknesses. Shielding materials used in place of the specified concrete provide the equivalent attenuation thickness prescribed for shielding gamma sources. In the case of shielding neutron sources, as-built shielding demonstrates equivalent attenuation to the barrier thicknesses in Table 12.3-5 for the NPM sources. Alternate shield material attenuation is to be demonstrated by achieving the radiation zones depicted in Figure 12.3-1a through Figure 12.3-2c. Alternate shielding is also to be verified to maintain compliance with 10 CFR 50.49, GDC 4, PDC 19, GDC 61, 10 CFR 50.34(f)(2)(vii), and other relevant requirements.The design provides equivalent density thicknesses for the barrier described using a variety of structural design solutions.
The reactor shielding calculations consider dose rates from fission neutrons, fission photons, and gamma output from buildup of radioisotopes in the reactor coolant. The NuScale Power Module (NPM) model is conservatively developed using methods similar to the building evaluations.
The fission neutron and fission photon output is based on a total power output of 250 MWt and energy spectra are described in Section 12.2. The gamma output from the reactor coolant is based on the reactor coolant isotopic inventory described in Section 12.2. In order to reduce complexity, some region densities (e.g., water and piping in the SGs) are homogenized in the MCNP model. This simplification does not result in significant differences in dose rates.
NuScale Final Safety Analysis Report Radiation Protection Design Features NuScale US460 SDAA 12.3-33 Draft Revision 2 RAI 12.3-1 Table 12.3-5: Reactor Building Shield Wall Geometry Elevation (Note 1)
Room No.
(Note 1)
Room Name (Note 1)
Radioactive Source North Wall East Wall South Wall West Wall Floor Ceiling 25-0 006 Module 01 CVC ion exchanger valve room CVC IXs, CVC FLTs 20 concrete 20 concrete 20 concrete 20 concrete Basemat 27.5 concrete 25-0 007 Module 01 CVC ion exchanger enclosure CVC IXs, CVC FLTs 20 concrete 20 concrete 20 concrete 20 concrete Basemat 24 concrete 25-0 008 Module 01 CVC filter enclosure CVC IXs, CVC FLTs 20 concrete 20 concrete 20 concrete 48 concrete Basemat 24 concrete 25-0 009 Module 02 CVC ion exchanger valve room CVC IXs, CVC FLTs 20 concrete 20 concrete 20 concrete 20 concrete Basemat 27.5 concrete 25-0 010 Module 02 CVC ion exchanger enclosure CVC IXs, CVC FLTs 20 concrete 20 concrete 20 concrete 20 concrete Basemat 24 concrete 25-0 011 Module 02 CVC filter enclosure CVC IXs, CVC FLTs 20 concrete 20 concrete 20 concrete 20 concrete Basemat 24 concrete 25-0 012 Module 03 CVC ion exchanger valve room CVC IXs, CVC FLTs 20 concrete 48 concrete 20 concrete 20 concrete Basemat 27.5 concrete 25-0 013 Module 03 CVC ion exchanger enclosure CVC IXs, CVC FLTs 20 concrete 48 concrete 20 concrete 20 concrete Basemat 24 concrete 25-0 014 Module 03 CVC filter enclosure CVC IXs, CVC FLTs 20 concrete 20 concrete 20 concrete 20 concrete Basemat 24 concrete 25-0 024 Module 04 CVC ion exchanger valve room CVC IXs, CVC FLTs 20 concrete 20 concrete 20 concrete 20 concrete Basemat 27.5 concrete 25-0 025 Module 04 CVC ion exchanger enclosure CVC IXs, CVC FLTs 20 concrete 20 concrete 20 concrete 20 concrete Basemat 24 concrete 25-0 026 Module 04 CVC filter enclosure CVC IXs, CVC FLTs 20 concrete 48 concrete 20 concrete 20 concrete Basemat 24 concrete 25-0 027 Module 05 CVC ion exchanger valve room CVC IXs, CVC FLTs 20 concrete 20 concrete 20 concrete 20 concrete Basemat 27.5 concrete
NuScale Final Safety Analysis Report Radiation Protection Design Features NuScale US460 SDAA 12.3-34 Draft Revision 2 25-0 028 Module 05 CVC ion exchanger enclosure CVC IXs, CVC FLTs 20 concrete 20 concrete 20 concrete 20 concrete Basemat 24 concrete 25-0 029 Module 05 CVC filter enclosure CVC IXs, CVC FLTs 20 concrete 20 concrete 20 concrete 20 concrete Basemat 24 concrete 25-0 030 Module 06 CVC ion exchanger valve room CVC IXs, CVC FLTs 20 concrete 20 concrete 20 concrete 48 concrete Basemat 27.5 concrete 25-0 031 Module 06 CVC ion exchanger enclosure CVC IXs, CVC FLTs 20 concrete 20 concrete 20 concrete 48 concrete Basemat 24 concrete 25-0 032 Module 06 CVC filter enclosure CVC IXs, CVC FLTs 20 concrete 20 concrete 20 concrete 20 concrete Basemat 24 concrete 25-0 034 LRW degasifier tank B room LRW DGS B 48 concrete 32 concrete 32 concrete 32 concrete Basemat 32 concrete 25-0 037 LRW degasifier tank A room LRW DGS A 48 concrete 32 concrete 32 concrete 48 concrete Basemat 32 concrete 26-0 040 Dry dock Pool water 48 concrete 48 concrete 48 concrete Basemat 26-0 to 100-0 041 Spent fuel pool Spent fuel assemblies, Pool water 48 concrete 48 concrete 48 concrete Basemat 36 concrete 26-0 to 100-0 042 Reactor pool Pool water 48 concrete 48 concrete 48 concrete 48 concrete Basemat 30 concrete 26-0 to 126-0 Module bays -
Modules 01-03 (gamma):
RXMNPMs, Pool water; (neutron):
NPMs (gamma): 51.75 concrete (TYP),
72 concrete (below EL 43-0); (neutron):
46.5" concrete (gamma): 51.75 concrete (TYP), 72 concrete (East pool wall below EL 43-0);
(neutron): 46.5" concrete 3.5 HDPE panels, 5% boron content (vertical bioshield)
(gamma):
51.75 concrete; (neutron): 46.5" concrete Basemat (gamma): 29 concrete; (neutron): 22" concrete 26-0 to 126-0 Module bays -
Modules 04-06 (gamma):
- RXMNPMs, Pool water; (neutron): NPMs 3.5 HDPE panels, 5% boron content (vertical bioshield)
(gamma): 51.75 concrete (TYP), 72 concrete (East pool wall below EL 43-0);
(neutron): 46.5" concrete (gamma): 51.75 concrete (TYP),
72 concrete (below EL 43-0);
(neutron): 46.5" concrete (gamma):
51.75 concrete; (neutron): 46.5" concrete Basemat (gamma): 29 concrete; (neutron): 22" concrete Table 12.3-5: Reactor Building Shield Wall Geometry (Continued)
Elevation (Note 1)
Room No.
(Note 1)
Room Name (Note 1)
Radioactive Source North Wall East Wall South Wall West Wall Floor Ceiling
NuScale Final Safety Analysis Report Radiation Protection Design Features NuScale US460 SDAA 12.3-37 Draft Revision 2 85-0 Pipe chases -
Modules 04-06 CVC discharge lines CVC discharge lines 48 concrete 16 concrete 16 concrete 16 concrete 100-0 Pipe chases -
Modules 01-03 CVC discharge lines CVC discharge lines 16 concrete 16 concrete 48 concrete 16 concrete 16 concrete 100-0 Pipe chases -
Modules 04-06 CVC discharge lines CVC discharge lines 48 concrete 16 concrete 16 concrete 16 concrete 16 concrete Note 1: Figure 1.2-8 through Figure 1.2-18 depict room locations.
Note 2: A 1 steel plate is placed above the entrance to the CVCS demineralizer valve gallery.
Note 3: The vertical pipe chase enclosure containing the CVCS resin transfer line the to the SRWS extends from EL 25-0 to EL 81-0.
Note 4: The reactor pool walls are typically 48" concrete, with the exception of areas immediately north, south, and east of the operating RXMNPMs, which are described as Module bays with theirthier own entry in the table.
Note 5: The vertical pipe chase enclosures containing the CVCS discharge lines from the RXMNPMs to the CVC heat exchangers extend from EL 70-0 to EL 111-7.
Note 6: Deviations from the wall thicknesses provided in this table are evaluated for impact to radiation shielding attenuation.
Table 12.3-5: Reactor Building Shield Wall Geometry (Continued)
Elevation (Note 1)
Room No.
(Note 1)
Room Name (Note 1)
Radioactive Source North Wall East Wall South Wall West Wall Floor Ceiling
NuScale Final Safety Analysis Report Radiation Protection Design Features NuScale US460 SDAA 12.3-38 Draft Revision 2 RAI 12.3-1 Table 12.3-6: Radioactive Waste Building Shield Wall Geometry Room Number (Note 1)
Source Term North Wall East Wall South Wall West Wall Floor Ceiling Labryinth Walls 005 LRW Processing Equipment Exterior subgrade wall Exterior subgrade wall 24 Concrete 24 Concrete Facility Basemat 24 Concrete
- 24 Concrete (Entrance)
- 24 Concrete (Drum dryer Area)
- 24 Concrete (Drum dryer Accumulator Tank Area)
- 24 Concrete (Ion Exchanger area)
- 24 Concrete (Charcoal Filter Area) 008 LRW LCW Sample Tank Pumps 24 Concrete 24 Concrete 24 Concrete 24 Concrete Facility Basemat 24 Concrete
- 24 Concrete (Doorway)
- 12 Concrete (Pump Divider) 009 LRW HCW Sample Tank Pumps 24 Concrete 24 Concrete 24 Concrete 24 Concrete Facility Basemat 24 Concrete
- 24 Concrete (Doorway)
- 12 Concrete (Pump Divider) 010 SRW PST Pumps 24 Concrete 36 Concrete 36 Concrete 24 Concrete Facility Basemat 24 Concrete
- 24 Concrete (Doorway)
- 12 Concrete (Pump Divider) 024 LRW LCW Collection Tank Pumps 24 Concrete 24 Concrete 24 Concrete 24 Concrete Facility Basemat 24 Concrete
- 24 Concrete (Doorway)
- 12 Concrete (Pump Divider) 025 LRW HCW Collection Tank Pumps 24 Concrete 24 Concrete 24 Concrete 24 Concrete Facility Basemat 24 Concrete
- 24 Concrete (Doorway)
- 12 Concrete (Pump Divider) 026 SRW SRST Pumps 36 Concrete 36 Concrete 24 Concrete 24 Concrete Facility Basemat 24 Concrete
- 24 Concrete (Doorway)
- 12 Concrete (Pump Divider) 011 LRW LCW Sample Tank 24 Concrete 24 Concrete 24 Concrete 24 Concrete Facility Basemat 24 Concrete N/A 012 LRW LCW Sample Tank 24 Concrete 24 Concrete 24 Concrete 24 Concrete Facility Basemat 24 Concrete N/A 018 LRW LCW Collection Tank 24 Concrete 24 Concrete 24 Concrete 24 Concrete Facility Basemat 24 Concrete N/A 019 LRW LCW Collection Tank 24 Concrete 24 Concrete 24 Concrete 24 Concrete Facility Basemat 24 Concrete N/A 013 LRW HCW Sample Tank 24 Concrete 24 Concrete 24 Concrete 24 Concrete Facility Basemat 24 Concrete N/A
NuScale Final Safety Analysis Report Radiation Protection Design Features NuScale US460 SDAA 12.3-39 Draft Revision 2 014 LRW HCW Sample Tank 24 Concrete 36 Concrete 24 Concrete 24 Concrete Facility Basemat 24 Concrete N/A 020 LRW HCW Collection Tank 24 Concrete 24 Concrete 24 Concrete 24 Concrete Facility Basemat 24 Concrete N/A 021 LRW HCW Collection Tank 24 Concrete 36 Concrete 24 Concrete 24 Concrete Facility Basemat 24 Concrete N/A 015 SRW PST 36 Concrete 36 Concrete 36 Concrete 36 Concrete Facility Basemat 36 Concrete N/A 016 SRW PST 36 Concrete 36 Concrete 36 Concrete 36 Concrete Facility Basemat 36 Concrete N/A 022 SRW SRST 36 Concrete 36 Concrete 36 Concrete 36 Concrete Facility Basemat 36 Concrete N/A 023 SRW SRST 36 Concrete 36 Concrete 36 Concrete 36 Concrete Facility Basemat 36 Concrete N/A 017 Pipe Chase 24 Concrete 24 Concrete 24 Concrete 24 Concrete Facility Basemat 24 Concrete N/A 029 PCU FLT 30 Concrete Exterior subgrade wall Exterior subgrade wall 30 Concrete Facility Basemat 30 Concrete N/A 030 PCU FLT 30 Concrete 30 Concrete Exterior subgrade wall 30 Concrete Facility Basemat 30 Concrete N/A 032 PCU IX 24 Concrete 30 Concrete Exterior subgrade wall 24 Concrete Facility Basemat 24 Concrete N/A 033 SRW Drum Storage 36 Concrete 36 Concrete 24 Concrete Exterior subgrade wall Facility Basemat 24 Concrete 20 Concrete 034 SRW HIC Storage 36 Concrete 36 Concrete 36 Concrete Exterior subgrade wall Facility Basemat 36 Concrete N/A 035 SRW HIC Filling 36 Concrete 36 Concrete 36 Concrete Exterior subgrade wall Facility Basemat 36 Concrete N/A 037 GRW Vapor Condensers / Gas Coolers 24 Concrete 36 Concrete 24 Concrete Exterior subgrade wall Facility Basemat 24 Concrete N/A 038 GRW Charcoal Beds 24 Concrete 24 Concrete 24 Concrete Exterior subgrade wall Facility Basemat 24 Concrete 24 Concrete Note 1: Refer to Figure 1.2-22 through Figure 1.2-24 for room locations.
Note 2. The equivalent attenuation to an additional 4.5 inches of lead is provided for a HIC process shield.
Note 3: The equivalent attenuation to an additional one inch of steel in addition to the LRWS process skid.
Note 4: A penetration to the HIC storage room is modeled with a 1ft concrete shadow shield.
Note 5: Deviations from the wall thicknesses provided in this table are evaluated for impact to radiation shielding attenuation.
Table 12.3-6: Radioactive Waste Building Shield Wall Geometry (Continued)
Room Number (Note 1)
Source Term North Wall East Wall South Wall West Wall Floor Ceiling Labryinth Walls