ML24037A134

LLC Response to NRC Request for Additional Information (RAI No. 10101) on the NuScale Standard Design Approval Application

ML24037A134
Person / Time
Site:	05200050
Issue date:	02/06/2024
From:	Shaver M NuScale
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RAIO-157045
Download: ML24037A134 (1)

Text

RAIO-157045 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com February 06, 2024 Docket: 52-050 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Response to NRC Request for Additional Information (RAI No. 10101) on the NuScale Standard Design Approval Application

REFERENCE:

NRC Letter Advance Copy for Prop Review Chapter 12, dated October 19, 2023, RAI# 10101 The purpose of this letter is to provide NuScale's response to NRC Requests for Additional Information (RAI), RAI# 10101, noted in the References above. The responses to the individual RAI questions are provided in the attached Enclosures.

This letter contains NuScale's response to the following RAI Questions from NRC RAI# 10101:

12.3-1 12.3.4.2-1 This letter makes no new regulatory commitments and no revisions to any existing regulatory commitments.

Please contact Jim Osborn at 541-360-0693 or at josborn@nuscalepower.com if you have any questions.

Sincerely, Mark Shaver Director, Regulatory Affairs NuScale Power, LLC Distribution:

Getachew Tesfaye, NRC Alina Schiller, NRC Mahmoud Jardaneh, NRC : NuScale Response to NRC Request for Additional Information RAI# 10101

RAIO-157045 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com :

NuScale Response to NRC Request for Additional Information eRAI No. 10101

NuScale Nonproprietary NuScale Nonproprietary Response to Request for Additional Information Docket: 052000050 RAI No.: 10101 Date of RAI Issue:10/19/2023 NRC Question No.: 12.3-1 Regulatory Basis 10 CFR Part 50, Appendix A, General Design Criteria (GDC) 61 requires in part that fuel storage and handling, radioactive waste, and other systems shall be designed with suitable shielding for radiation protection.

10 CFR 52.137(a)(5) requires that an application for an SDA must include information on the kinds and quantities of radioactive materials expected to be produced in operation and the means for controlling and limiting radioactive effluents and radiation exposures within the limits set forth in part 20.

Issue FSAR Section 12.3.2.3 indicates that the credited radiation shielding barriers for the reactor building and radioactive waste building are provided in terms of nominal concrete equivalent thicknesses and that the design provides equivalent density thicknesses for the barrier described using a variety of structural design solutions. During the audit, NuScale described the approach of using equivalent density thickness as using the ratio of densities of the materials to determine the thickness of the replacement material (for example, if a material has a density 2.25 times that of the concrete shield thickness identified in Tables 12.3-5 or 12.3-6 in FSAR Chapter 12 of the NuScale SDAA, the replacement material thickness will be 2.25 times less thick than the provided concrete value). However, the amount of radiation attenuated by a shield is not only dependent on the ratios of density and thickness. Even for gamma radiation, mass attenuation coefficients are different for different materials. Determining appropriate radiation shielding is even more complex for neutron radiation. Therefore, if a replacement radiation shielding material is used, the replacement shielding material should provide at least equivalent radiation attenuation as the specified material.

Furthermore, NuScale SDAA, Part 8, Table 3.11-1, item 4 and Table 3.12-1, item 1 provide the Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) for the reactor building and radioactive waste building radiation shielding barriers. The acceptance criteria for these ITAAC

NuScale Nonproprietary NuScale Nonproprietary include that a report exists and concludes that the radiation attenuation capability of the radiation shielding barriers is greater than or equal to the required attenuation capability of the approved design (note that Table 3.11-1, item 5 also includes a similar ITAAC for reactor building radiation shield doors). As a result, the equivalent density thickness approach described in FSAR Section 12.3.2.3 appears inconsistent with the ITAAC acceptance criteria for these ITAAC.

Information Requested Provide additional information describing how the alternative radiation shielding approach of using equivalent density thicknesses, as mentioned in FSAR Section 12.3.2.3 and as described during the audit, is acceptable. Include information on how radiation protection of workers and equipment will be ensured using the specified approach and discuss potential impacts on designated plant radiation zones and equipment qualification radiation zones and total integrated dose specifications in the FSAR.

Also, discuss the apparent discrepancy between the equivalent density thicknesses approach and the ITAAC acceptance criteria for ITAAC, item 4 in Table 3.11-1 and item 1 in Table 3.12-1 of SDAA, Part 8 and how the ITAAC will be addressed if alternative radiation shielding is used.

As appropriate, update the SDAA to incorporate any updated information or to address any inconsistencies.

NuScale Response:

Tables 12.3-5 and 12.3-6 of the Standard Design Approval Application (SDAA) describe the Radioactive Waste Building and Reactor Building shielding barriers in terms of minimum nominal concrete thicknesses. Although a few shielding barriers in Table 12.3-5 are exposed to neutron radiation, the overall dose contribution of the attenuated neutron field in these areas is a negligible fraction in comparison to the attenuated gamma field. Below the ultimate heat sink pool level, neutron contributions to overall dose rates after attenuation by the pool water and module bay walls are negligible compared to the gamma contributions. Above the ultimate heat sink pool level, neutrons are a significant contributor to overall dose rates; however, dose rates beyond the module bay walls are primarily dominated by indirect streaming paths rather than the attenuation functions of the shield barrier. For the shield barriers surrounding the spent fuel rack in the spent fuel pool, the neutron dose contributions from the spent fuel are sufficiently attenuated by the pool water prior to interacting with the shield barriers. Therefore, overall dose rates in the Reactor Building gallery space beyond the shield barriers are dominated by gamma

NuScale Nonproprietary NuScale Nonproprietary radiation with negligible neutron contributions. Shield barriers described in Table 12.3-6 are not exposed to neutron radiation within the Radioactive Waste Building.

The use of nominal concrete thickness in SDAA Tables 12.3-5 and 12.3-6 reflects the shielding required to protect personnel and equipment from the effects of the predominantly gamma radiation field. The neutron shield values of Table 12.3-5 include shielding for gamma radiation produced as a result of neutron radiation. The shield barrier thicknesses also include conservatism to ensure the adequacy of the resultant radiation zones. The fission neutron source strength defined in SDAA Table 12.2-1 applies a uniform axial peaking factor based on the limiting radial power profile identified in SDAA Figure 4.3-8 to bound dose rates. Nominal concrete shielding thickness is therefore representative of the attenuation thickness required to achieve the radiation zones of SDAA Figures 12.3-1a through 12.3-2c.

Inspections, tests, analyses, and acceptance criteria (ITAAC) 03.11.04 and 03.12.01, respectively, verify radiation shielding barriers in the RXB and RWB have the required attenuation capability, as specified in the SDAA. The respective ITAAC discussions further clarify these ITAAC. As described, an ITAAC inspection is performed to verify wall materials and thicknesses, and these shielding barriers must meet or exceed the attenuation capability of the materials at the thicknesses specified in SDAA FSAR Tables 12.3-5 (RXB) and 12.3-6 (RWB).

The last sentence of the discussion states, Attenuation capabilities are determined based on wall materials and thicknesses, and an analysis and report will conclude that attenuation capabilities are greater than or equal to the approved design. The attenuation capabilities established in the approved design are those listed in Tables 12.3-5 and 12.3-6. Therefore, if a downstream licensee were to apply an equivalent density thickness, the ITAAC report would include reconciliation between the attenuation capability of the SDAA and the capability in the as-built plant. The SDAA is revised, below, to correct the table references in these ITAAC discussions.

COL Item 3.11-2 is revised to include mild environment equipment whose qualification basis may be impacted by changes to environmental service conditions including changes to radiological shielding.

Impact on US460 SDA:

FSAR Section 12.3, 3.11, Part 8, and Table 1.8-1 has been revised as described in the response above and as shown in the markup provided in this response.

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

License Conditions; ITAAC Shared Structures, Systems, and Components and Non-Structures, Systems, and Components Based Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Design Descriptions and ITAAC NuScale US460 SDAA 148 Draft Revision 2 RAI 12.3-1 03.11.04 FSAR Section 12.3, Radiation Protection Design Features, provides the design bases for radiation shielding, including type, form and material properties utilized in specific locations. Radiation shielding is provided to meet the radiation zone and access requirements for normal operation and post-accident conditions, and to demonstrate compliance with 10 CFR 50.49, GDC 4, PDC 19, GDC 61, 10 CFR 50.34(f)(2)(vii), and other relevant requirements. Compartment walls, ceilings, and floors, or other barriers provide shielding.

An ITAAC inspection is performed of the RXB radiation barriers to verify wall materials and thicknesses.

The required thicknesses are specified in FSAR Table 12.3-56. Attenuation capabilities are determined based on wall materials and thicknesses, and an analysis and report will conclude that attenuation capabilities are greater than or equal to the approved design.

03.11.05 Not UsedFSAR Section 12.3.2 provides the design bases for radiation shielding. Radiation shielding is provided to meet the radiation zone requirements for normal operation and control room access requirements for post-accident conditions. Radiation attenuating doors must meet or exceed the radiation attenuation capability of the wall within which they are installed.

An ITAAC inspection is performed to verify that the RXB radiation attenuating doors are installed in their design location and have a radiation attenuation capability that meets or exceeds that of the wall within which they are installed in accordance with the approved door schedule design.

03.11.06 FSAR Section 3.8.4 and Appendix 3B provide descriptive information, including plans and sections of each Seismic Category I structure, to establish that there is sufficient information to define the primary structural aspects and elements relied upon for the structure to perform the intended safety functions. Critical dimensions are identified in FSAR Appendix 3B. The RXB and its design basis loads are discussed in FSAR Section 3.8.4. Critical sections are the subcomponents of individual Seismic Category I structures (i.e., shear walls, floor slabs and roofs, structure-to-structure connections) that are analytically representative of an essentially complete design. Design basis load combinations are shown in FSAR Tables 3.8.4-1 and 3.8.4-2.

A reconciliation analysis of the as-built RXB is performed to ensure the RXB maintains its structural integrity in accordance with the approved design under the actual design basis loads, and the in-structure responses for the RXB are enveloped by those in the approved design. The design summary report provides criteria for the reconciliation between design and as-built conditions, as described in FSAR Section 3.8.4.

03.11.07 FSAR Section 3.2.1, Seismic Classification, discusses that per RG 1.29, some SSC that perform no safety-related functions could, if they failed under seismic loading, prevent or reduce the functioning of Seismic Category I SSC.

An ITAAC inspection and analysis is performed to verify that the as-built non-Seismic Category I SSC where there is a potential for adverse interaction with the RXB or a Seismic Category I SSC in the RXB will not impair the ability of Seismic Category I SSC to perform their safety functions as demonstrated by one or more of the following criteria:

i. Seismic Category I SSC are isolated from non-Seismic Category I SSC so that interaction does not occur.

ii. Seismic Category I SSC are analyzed to confirm that the ability to perform their safety functions is not impaired as a result of impact from non-Seismic Category I SSC.

A non-Seismic Category I restraint system designed to Seismic Category I requirements is used to ensure that no interaction occurs between Seismic Category I SSC and non-Seismic Category I SSC.

Table 3.11-2: Reactor Building Inspections, Tests, Analyses, and Acceptance Criteria Additional Information(1) (Continued)

ITAAC No.

Discussion

License Conditions; ITAAC Shared Structures, Systems, and Components and Non-Structures, Systems, and Components Based Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Design Descriptions and ITAAC NuScale US460 SDAA 153 Draft Revision 2 Audit Question A-Part 8-3.12-4 Table 3.12-2: Radioactive Waste Building Inspections, Tests, Analyses, and Acceptance Criteria Additional Information(1)

ITAAC No.

Discussion RAI 12.3-1 03.12.01 FSAR Section 12.3, Radiation Protection Design Features, provides the design bases for radiation shielding, including type, form and material properties utilized in specific locations. Radiation shielding is provided to meet the radiation zone requirement for normal operation and post-accident conditions and to demonstrate conformance with GDC 61, RG 4.21, RG 8.8, and other relevant requirements. Compartment walls, ceilings, and floors, or other barriers provide shielding.

An ITAAC inspection is performed of the RWB radiation barriers to verify wall materials and thicknesses.

The required thicknesses are specified in FSAR Table 12.3-67. Attenuation capabilities are determined based on wall materials and thicknesses, and an analysis and report will conclude that attenuation capabilities are greater than or equal to the approved design.

03.12.02 FSAR Section 12.3 provides the design bases for radiation shielding. Radiation shielding is provided to meet the radiation zone requirements for normal operation and post-accident conditions, and to demonstrate conformance to RG 4.21 and RG 8.8. Radiation attenuating doors must meet or exceed the radiation attenuation capability of the wall within which they are installed.

An ITAAC inspection is performed to verify that the RWB radiation attenuating doors are installed in their design location and have a radiation attenuation capability that meets or exceeds that of the wall within which they are installed in accordance with the approved door schedule design.

03.12.03 The RW-IIa RWB and its design basis loads are discussed in FSAR Section 3.8.4. Design basis loads for RW-IIa structures asare listed in RG 1.143.

An inspection and analysis are performed for the RW-IIa portions of the as-built RWB. For the RW-IIa portions, a report exists and concludes the deviations between the drawings used for construction are reconciled, and these portions of the RWB maintain structural integrity under the design basis loads.A reconciliation analysis of the as-built RW-IIa RWB is performed to ensure that deviations between the drawings used for construction and the as-built RW-IIa RWB are reconciled and the RW-IIa RWB maintains its structural integrity under the design basis loads in accordance with the approved design under the actual design basis loads, and the in-structure responses for the RWB are enveloped by those in the approved design.

Note:

1) References to Tables and Figures refer to ITAAC unless the reference specifically states FSAR Tables or Figures.

NuScale Nonproprietary NuScale Nonproprietary Response to Request for Additional Information Docket: 052000050 RAI No.: 10101 Date of RAI Issue:10/19/2023 NRC Question No.: 12.3.4.2-1 Regulatory Basis 10 CFR 50.34(f)(2)(xix) requires that instrumentation be provided for adequate monitoring of plant conditions following an accident that includes core damage.

10 CFR 50.34(f)(2)(xvii), item D requires, in part, that instrumentation is provided to measure, record and readout containment radiation intensity (high level) in the control room.

10 CFR Part 50, Appendix A, General Design Criterion 64 requires that means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents.

Issue NuScale SDA Section 12.3.4.2 states that the fixed area radiation monitors used for post-accident monitoring (PAM) have ranges that consider the maximum calculated accident levels and are designed to operate effectively under the environmental conditions caused by an accident. It also states that the PAM monitors conform to Regulatory Guide (RG) 1.97, Revision 5, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants.

The staff notes that the under-the-bioshield radiation monitors in the NuScale design serve as the alternative to the containment high range monitors in large light water reactors (LWRs).

NUREG-0737, Clarification of TMI Action Plan Requirements, Table II.F.1-3 specifies that the containment high range radiation monitors shall have the capability to detect and measure the radiation levels within the reactor containment during and following the accident. NUREG-0737 also references RG 1.97, Revision 2, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident. Specifically, RG 1.97, Revision 2, states that one of the purposes of the containment high range radiation monitors is long-term surveillance. In large LWRs, the containment high range radiation monitors are typically qualified to function long term following a core damage accident and are

NuScale Nonproprietary NuScale Nonproprietary used to aid in accident diagnosis and control and are used in emergency classifications.

The NuScale under-the-bioshield radiation monitors are designated as Type B, C, and F post-accident monitoring variables in the NuScale SDAA design. RG 1.97, Revision 5 references Institute of Electrical and Electronics Engineers (IEEE) Standard (Std.) 497-2016, IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations, which specifies that for Type C variables, the required operating time shall be the duration for which the measured variable is required by the plants licensing basis document (LBD) or at least 100 days following the start of an accident. However, SDA Table 19.2-8 specifies an equipment survivability duration in Table 19.2-8 of only 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after core damage.

The staff notes that the radiological core damage conditions under the bioshield are generally not more severe than the core damage conditions inside containment in large LWRs and it does not appear that other environmental conditions such as temperatures or pressures are more severe either.

In addition, in exemption 16, NuScale requests an exemption from taking post-accident sampling. Post-accident sampling requirements are in place as a means to provide data in the post-core accident environment. With the under-the-bioshield radiation monitors only surviving for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and no post-accident sampling, it appears that there may be no means to determine the radiological conditions inside containment or under the bioshield beyond 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Furthermore, there doesnt appear to be any radiation monitoring equipment anywhere in the facility designated to operate more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following a core damage accident.

SDA Section 12.3.4.2 provides information about the under-the-bioshield monitors; however, the SDA doesnt provide much detail for its specific uses in accident conditions, beyond that the monitors are intended to detect fuel damage.

Information Requested Provide additional information on the intended uses of the NuScale under-the-bioshield radiation monitors, following a core damage accident. It is the staffs understanding that the radiation levels under the bioshield could be thousands or more R/hr after the first 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following the start of the core damage accident. Please provide justification as to why it is acceptable for the information regarding the radiological conditions under the bioshield to potentially be unavailable after the first 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for the plant staff to take appropriate actions, provided the information in IEEE 497-2016, especially considering accidents that may not progress as anticipated.

NuScale Nonproprietary NuScale Nonproprietary Additional Information Requested:

1. Can NuScale provide any additional information regarding how the monitors are adequate to address potential uncertainties in accident progression (or if it is not necessary to be concerned about uncertainties in accident progression for the monitors, why that would be the case)?

2. Could NuScale please clarify the last sentence of the response. The staff dont see the referenced license condition in Part 8 of the application or a COL item or commitment in the application indicating that the COL applicant will provide a license condition. It is also unclear what report is being provided in the COL application.

NuScale Response:

The function of the area radiation monitor (ARM) under the bioshield is to provide indication of containment gamma radiation under normal conditions and to detect the onset of core damage, the extent of core damage, and failures of containment integrity, under accident conditions.

Section 12.3.4.2 of the Standard Design Approval Application (SDAA) states the radiation monitors under the bioshield are environmentally qualified to survive an accident and perform their design functions. In accordance with SDAA Table 3.11-1, the environmentally qualified operating time for the under the bioshield ARMs under a harsh environment is 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> (30 days) after initiation of a design basis accident.

Table 19.2-8 of the SDAA, Equipment Survivability List, defines a 48-hour period (after core damage) of under the bioshield ARM equipment survivability for severe accidents. Equipment survivability in a radiation environment compares the severe accident dose based on core damage to the environmental qualification design basis dose. As required by SDAA Section 19.2.3.3.8, Equipment Survivability, cases where the severe accident dose is larger are subject to qualitative assessments, testing, or additional analysis to ensure equipment survivability.

Subsequently, SDAA Revision 2 COL Item 19.2-4 ensures licensees use the most limiting equipment dose required in the procurement specification for any equipment defined in Table 19.2-8.

As a means of providing defense-in-depth, each NuScale Power Module has two bioshield radiation monitors that are Type B, C, and F PAM variables. If the bioshield radiation monitors become inoperable, additional (Type E PAM variable) radiation monitors are available to provide assessment of the radiological event. The additional monitors include, but are not limited to the following:

Refuel Pool Area Radiation Monitor

NuScale Nonproprietary NuScale Nonproprietary

Refuel Pool Area Continuous Airborne Radiation Monitor

North and South process sampling system sample panel area radiation monitors

North and South Steam Gallery Continuous Airborne Radiation Monitor

Reactor Building Plant Exhaust Stack continuous airborne radiation monitor These monitors provide assurances that accident progressions are effectively monitored by plant operators utilizing the design features of the plant.

As defined in SDAA Section 19.2.3.3.8, post-accident monitoring is only relied upon to provide information on severe accident conditions as required by 10 CFR 50.34(f)(2)(xix). The ARM under the bioshield is not relied upon for mitigation of the severe accident. After the equipment survivability period of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, plant conditions are stable and the hypothetical release from containment is either steadily decreasing or terminated completely, such that changing atmospheric conditions are the relevant input to operational decision-making.

Impact on US460 SDA:

There are no impacts to US460 SDA as a result of this response.