ML24344A099

Enclosure Plm PSAR Readiness Assessment Observation Table

ML24344A099
Person / Time
Site:	99902117
Issue date:	12/19/2024
From:	NRC/NRR/DANU/UAL2
To:	Dow Chemical Co, X-Energy
Ondra Dukes, NRR/DANU
Shared Package
ML24344A095:ML24344!095	List: ML24344A099 ML24344A100
References
Download: ML24344A099 (1)

Text

Enclosure Project Long Mott Draft PSAR Readiness Assessment Observations The following definitions are used to categorize each observation:

Category A: Preliminary Safety Analysis Report (PSAR) Gap Information that the U.S. Nuclear Regulatory Commission (NRC) staff expects to be required to meet the information requirements in Title 10 of the Code of Federal Regulations (10 CFR) section 50.34(a) and it was not provided in the draft PSAR.

Category B: Items Requiring Additional Information Items that the NRC staff perceives as needing justification or additional information to support a regulatory finding or which could impact the schedule and resource estimate for the NRC staffs review.

Category C: Other Other observations that should be addressed or considered by Dow Chemical Company (Dow) and X Energy, LLC., (X-energy) to support the clarity or quality of the application. If unaddressed, together, they could negatively impact the NRC staffs review of the application, including resources and schedule.

ID Chapter Section Observation Category 1

General All applicable Information provided to the NRC staff included the use of placeholder items in (yellow highlight) and tables and figures (e.g., section 3.2) that are blank and are to be completed by the construction permit (CP) application. Many of these placeholder items are needed to meet the 10 CFR 50.34(a) information requirements.

A 2

3 3.2.3 Regarding the key results of AOOs, DBEs, and BDBEs, this section states that No LBEs meet the criteria for risk significance. Consistent with NEI 18-04, risk-significant LBEs are those with frequencies within 1% of the F-C target with site boundary doses exceeding 2.5 mrem.

NEI 18-04 states that LBEs are regarded as risk-significant if the combination of the upper bound (95% tile) estimates of the frequency and consequence of the LBE are within 1% of the frequency-consequence (F-C) Target and the upper bound 30-day TEDE dose at the EAB exceeds 2.5 mrem.

Clarify if X-energy is using the criterion in NEI 18-04 to determine risk-significant LBEs.

B

2 ID Chapter Section Observation Category 3

3 3.2.3.1 Section 3.2.3.1, Specified Acceptable Radionuclide Release Design Limits Results, states that Table 3.2-2 demonstrates that the AOOs meet the SARRDL, demonstrating compliance with PDCs 10, 11, and 12.

(1) In Revision 3 of the Xe-100 PDC topical report, PDC 11 subsumed PDC 12; therefore, PDC 12 should not be used.

(2) In the PDC topical report, there are other PDCs, e.g., PDC 26, that contain design criteria to meet SARRDL for AOOs. Explain why only PDC 10 and PDC 11 are listed in this section.

(3) Clarify how table 3.2-2, Bounded AOOs, DBEs, and BDBEs, which does not contain any information in the section 3.2, Licensing Basis Event Summary and Evaluation Methodology, file in the electronic reading room, would, on its own, demonstrate compliance with the PDCs listed. It appears that additional information in other sections of the PSAR would be needed for the demonstration.

C 4

3 3.2.5 Section 3.2.5, Conformance with Offsite Dose Limits, states that For the Xe-100 plant, the EAB and LPZ are identical: 400 meters from the corner of each NIAB.

It is the NRC staffs understanding that the NIABs are no longer part of the re-configured Xe-100 design. Clarify whether the statement above remains valid.

C 5

3 3.2.5 Section 3.2.5 states that X-energy has chosen Option 1 between the two options discussed in RG 1.253 to meet 10 CFR 50.34(a)(1)(ii)(D).

For Option 1, RG 1.253 discusses the potential need to include an exemption request from the regulations in 10 CFR 50.34 that require an assumed major accident to demonstrate containment performance and to confirm that the EAB and LPZ doses are below the reference values in the regulations. Clarify whether or not X-energy plans to request an exemption as discussed in the guidance.

B 6

3 3.7.1 Section 3.7.1, Plant Overview, refers to figure 1.1.3-2, Plant Layout, for the layout of the Xe-100 site. Figure 1.1.3-2 is not available as part of this readiness assessment. Figure 1.1.3-2, reflecting the latest design, should be included in the PSAR for the CPA. In addition, the descriptions of the relevant structures should be sufficiently detailed in chapters 1, 6, and 7 of the PSAR. This information would allow the NRC staff to understand the descriptions, layout, and functions of safety-significant structures in order to support an efficient NRC staff evaluation of relevant structures in accordance with 10 CFR 50.34(a).

B

3 ID Chapter Section Observation Category 7

3 3.7.2.1 (1)

Section 3.7.2.1, Seismic Design Basis for Safety Related Structures, states that the site-specific seismic hazard analysis and GMRS information for the site is not yet available and approximate input is adopted based on publicly available information developed in response to the Post-Fukushima NTTF recommendation 2.1 as well as:

(a) Adjustments to account for NGA-EAST amplifications in the 0.5 Hz to 5 Hz range and, (b) Scaling the spectral shape for the NGA-EAST adjusted GMRS with a peak ground acceleration (PGA) of 0.07g to the minimum PGA of 0.1g documented in 10 CFR Part 50 Appendix S(IV)(a)(1)(i).

Describe the planned approach to evaluate the GMRS following a site-specific seismic hazard and GMRS determination, and include the description in PSAR section 3.7.2.1.

B 8

3 3.7.2.1 (2)

The last paragraph of Section 3.7.2.1 refers to a median capacity of 2.46 times the SR DRS PGA and a composite variability of 0.86. The composite variability appears large as compared to the values in ASCE 43-19. Provide the steps for the calculation of the composite variability of 0.86.

In addition, explain how the calculation for the median and the composite variability would change when the design limit state is not limit state D in ASCE 43-19, with reference to the approach in commentary section C1.5 of ASCE 43-19 which shows calculations for limit state D only.

Section 4.2.3, Structural Fragility, of NRCs RIL 2021-04 (ML21113A066) has examples of these evaluations when using generic ASCE 43-19 fragilities.

B 9

3 3.7.2.2 Section 3.7.2.2, Loads, states that at the current state of the design, jet impingement, missile impact, and pipe reactions loads generated by a postulated pipe break are not considered in the load combinations because their effects are expected to be local. Local effects under some circumstances may lead to deformations of the whole structure, for example, as part of local collapse mechanisms.

Confirm that those loads will be considered in the design and revise PSAR Section 3.7.2.2 accordingly.

C 10 3

3.7.2.2.8 Section 3.7.2.2.8, Other Loads, states that thermal loads, accident pressure, pipe reaction and pipe break are judged not to control the structural member sizes and are not considered in the current stage of design (CP stage). Explain the basis for this assertion in PSAR Section 3.7.2.2.8.

B

4 ID Chapter Section Observation Category 11 3

3.7.2.3 Section 3.7.2, Load Combinations, states that load combinations will be per ACI 349 as modified in accordance with table 1, Loads, Load Factors and Load Combinations, of RG 1.142. It also states that structural steel elements in the SR RXSS are to be designed using the load combinations from AISC N690-18 Section NB2.5, Load and Resistance Factor Design (LRFD), and load factors corresponding to the acceptance criteria applied for the design per AISC N690. The NRC staff endorsed AISC N690-18 in RG 1.243 with staff positions that modified a few load factors for the load combinations in AISC N690-18. X-energy should justify the approach used.

B 12 3

3.7.2.4.1 (1)

Section 3.7.2.4.1, Use of Computer Codes, states that the RXSS is modeled and analyzed using the provisions of ASCE/SE 4-16 and the computer program SAP2000 version 22.1.0, to calculate forces and moments in the frame elements and at boundary conditions. Explain how SAP2000 is used in conjunction with the results from the soil-structure interaction (SSI) analysis.

B 13 3

3.7.2.4.1 (2)

Section 3.7.2.4.1 states that the SSI analysis of the SR Foundation is performed using the LS-DYNA FEA code using the effective seismic input method (ESIM). It also states that the ESIM method is used to incorporate forces in the SSI model using only the free field ground motion at the soil structure interface.

1. Describe how the ESIM model is used for deep foundations using piers and a control point for the ground motion at the surface.

2. Describe how ground motions in various depths are used to account for the provisions in section 2.3, Method to Define Design Response Spectra at Various Depths in the Site Profile, of ASCE 43-19.

B 14 3

3.7.2.4.2 Section 3.7.2.4.2, Seismic Design Parameters, states that for the DRS defined in section 3.7.2.1, the grade elevation is the control point location to compare the DRS to a site-specific GMRS and for the dynamic analysis of the RXSSs and SR Foundation.

Explain what considerations were used to choose the grade elevation for the control point and not, for example, the elevation at the bottom of an excavation and backfilling for the foundation basemat.

B

5 ID Chapter Section Observation Category 15 3

3.7.2.4.2 and 3.7.2.4.3 Section 3.7.2.4.2 states that nonlinear time history analysis has not been performed at this stage and that the final RIPB structural design criteria will provide the basis for the nonlinear time history analysis methodology discussed in section 3.7.2.4.3, Analysis Methods, and will address the applicable guidance in NUREG-0800 (Section 3.7). Section 3.7.2.4.3 provides an approach for the use of results from multiple time-history analyses. SRP section 3.7.2, Seismic System Analysis, which provides guidance to the NRC staff, does not contain detailed guidance on the review of inelastic/nonlinear analysis. Rather than referencing the SRP, the PSAR should clarify if the criteria for inelastic/nonlinear analysis and associated bases will be detailed in the FSAR.

C 16 3

3.7.2.4.3 Section 3.7.2.4.3 states that time history methods are used for the calculation of ISRS for the SR structures. Explain what approaches are used for the generation of time history methods.

B 17 3

3.7.2.4.5 Describe what structures, foundation basemats and deep foundation are included in the SSI analysis and model. Does the analysis include only the SR basemat, the SR structures on the SR basemat and the deep foundations for the SR basemat, or does it include the surrounding NSRST structures, basemat and deep foundations?

Also describe how the provisions in Section 2.3 of ASCE 43-19 will be used to define ground motions at various depths and their use in the SSI analysis.

The PSAR should discuss what structures and their foundations are included in the SSI analysis, including structures and foundations in close proximity to safety-related structures and foundations.

The PSAR should also indicate how the SSI analysis conforms to the provisions in ASCE 4-16, chapter 5, and in ASCE 43-19, Section 2.3.

B 18 3

3.7.2.5.2 Section 3.7.2.5.2, Safety Related Foundation and Piers Structural Analysis Results, addresses structural analysis results for the SR drilled pier foundation. It discusses the geotechnical load bearing capacity of the drilled piers as well as the structural design of the piers. This section refers to ACI 349-13 for the structural design of the piers.

1. This section should include the codes and standards or guidance for the geotechnical design of the piers, namely, for the calculation of their side friction and end bearing capacities.

2. This section should include a description of the planned approach used at the CP stage for the sizing of the piers and their spacing for geotechnical purposes.

B

6 ID Chapter Section Observation Category

3. It should include the general approach to address determining the needed seismic gaps to accommodate rotations and translations.

7 ID Chapter Section Observation Category 19 3

3.7.3.1 Section 3.7.3.1, Seismic Design Basis for NSRST Structures, states that the non-SR structures are designed following ASCE 7-16 provisions for risk category IV facilities with the MCER = 1.0xDBHL. It further states that to address the seismic interaction between non-SR and adjacent SR structures (referred as II/I seismic interaction) the MCER will be 1.5xDBHLT. It also states that sufficient separation is provided between the SR and non-SR structures at the seismic DBHL. Clarify if the seismic design loads in Section 3.7.3.1 apply to non-SR structures or only NSRST structures or both.

(a) Clarify if the statement in Section 3.7.3.1 which states that non-SR structures are designed following ASCE 7-16 provisions for Risk Category IV with MCER = 1.0xDBHL applies to all NSRST structures? Does it apply to non-SR structures other than NSRST as the first sentence in section 3.7.3.1 appears to say?

(b) Clarify if the separation is at the seismic DBHL for all NSRST structures or if it is 1.5xDBHL for those NSRST structures using the II/I seismic interaction.

(c) Clarify if the statement for a MCER=1.5xDBHL for II/I evaluations means that ASCE 7-16 will be used with this MCER, or they will be used with the MCER=1.0xDBHL but with a risk factor of 1.5. Also, describe how the 2/3 factor to MCER and response reduction factor, R, will be incorporated in developing seismic demands and loads.

(d) Section 3.7.3.1 does not provide an approach to assess the impact on the risk of a collapse of the NSRST structure on SR SSCs, and if the proposed approach will meet with the risk performance expected for the design. It is understood that a full seismic PRA is not in the scope of the CP.

However, section 3.7.3.1 does not provide a preliminary assessment of how risk performance can be assessed at this stage and how it will be evaluated as the design is finalized. For example, it does not say at what stages, CP or final, fragility analyses will be used to assess events in the BDBE region of a F-C curve. However, section 7.3.1.1.1, Shield Structure Functions and Design Criteria, states that as the design matures, the seismic demand, structural demand and structural strength and failure modes are properly accounted for in the seismic fragility and final probabilistic risk assessment to be detailed in the FSAR. The PSAR should describe a high-level process for the RIPB design, including how the design decisions will be made in an iterative process similar to the description presented in meetings with the NRC staff (ML24221A408).

(e) Are the DBHLs addressed in section 3.7.3.1 applicable to all NSRST structures identified in section 3.7.3.4, Structural Analyses of NSRST Structures, in addition to the SS? Specifically, are B

8 ID Chapter Section Observation Category they applicable to the SGSS (section 3.7.3.4.2, Steam Generator Support Structures (SGSS)

Structural Analysis), the shared non-safety related foundation and piers structural analysis (section 3.7.3.4.3, Shared Non-Safety related Foundation and Piers Structural Analysis), the FHAB (section 3.7.3.4.4, Fuel Handling Annex Building (FHAB) Structural Analysis)? The PSAR should clarify if the fragility analyses and their use in the PRA for all NSRST structures, in addition to the SS building, are to be detailed in the FSAR.

(f) Also clarify that section 3.7.3.4.3 refers to non-SR foundations and piers shared with NSRST foundations and piers. Clarify as well as if the foundation of the NSRST SGSS is also NSRST and that the foundation of the FHAB is also NSRST.

9 ID Chapter Section Observation Category 20 3

3.7.3.4.1 (1)

Section 3.7.3.4.1 states that the SS will be analyzed following applicable ASCE 7-16 sections. It also states that the SS is modeled and analyzed in a three-dimensional finite element model using the SAP2000 structural analysis program in accordance with ASCE 7-16. X-energy should clarify if the analysis uses the ASCE 7-16 option for time-history analyses or the standard analysis provisions in ASCE 7-16.

C 21 3

3.7.3.4.1 (2)

Section 3.7.3.4.1 states that the SS is evaluated to confirm that it meets the collapse prevention limit state outlined in ASCE 41. It also states that this analysis will be based on the SSE design response spectra. Is the SSE also the DBHL? Is this approach also used to evaluate the collapse of the SS at 1.5xDBHL as referred to in Section 3.7.3.1?

How would the evaluation using the approach of ASCE 41 at the SSE or at the 1.5xDBHL be used in the risk assessment or in the evaluation of a fragility for the SS for use in conjunction with risk considerations at the CP stage while taking into account the comments for sections 3.7.3.1 (c) and (d)?

Describe the approach that will be used to derive the fragility of this SS precast concrete structure for use in the SPRA and to capture the behavior for BDBEs.

B 22 3

3.7.3.4.2 Clarify if the SGSS is analyzed using the ASCE 7-16 options for the time-history analysis or the standard provisions in ASCE 7-16 and the three-dimensional model finite element model as well as SAP2000.

C 23 3

3.7.3.5.1 Section 3.7.3.5.1, Shield Structure (SS) Structural Analysis Results, states that the SS concrete member strengths are calculated in accordance with the strength design provisions of ACI 318. Since ACI 318 does not have provisions for the design of concrete structures against impactive or impulsive loads from tornado and hurricane missiles or accidental explosions, PSAR Section 3.7.3.5.1 should include information addressing these loads on the SS, including the codes and standards that will be used.

B 24 3

3.7.3.5.1 Section 3.7.3.5.1 refers to steel members and connection strengths for the SS that are calculated with the strength design provisions of AISC 360-16. The PSAR should describe what connections and steel members this statement refers to, including connections between precast modules.

B 25 3

3.7.4.6 Section 3.7.4.6, Steel-Concrete Composite (SC members), refers to SC members for SR structures.

The PSAR should describe which SR structures use SC members.

B

10 ID Chapter Section Observation Category 26 3

Table 3.7-1 (pg 14)

Table 3.7-1, Load Combinations for NSRST Concrete SSCs, contains the load combinations for NSRST concrete SSCs. Although these are NSRST structures, some of the load combinations include loads that are typical of nuclear power plant safety-related structures, for example; SSE, OBE, and tornado/hurricane wind. Were the modifications to the load combinations or factors in the ACI 349 load combinations in RG 1.142 considered in the load combinations in Table 3.7-1? X-energy should justify the chosen load combinations and factors considered.

The SS is an NSRST structure and one of its functions is to provide protection from hurricane and tornado missiles. Table 3.7-1 for the NSRST concrete structures should include loads from tornado or hurricane missiles given that the function of the SS is to shield the SR SSCs inside the SS from tornado and hurricane borne missiles.

C 27 3

Table 3.7-1 (pg 15)

Table 3.7-1, Load Combinations for NSRST Steel SSCs, has the load combinations for NSRST steel SSCs. Although these are NSRST structures, some of the load combinations include loads that are typical of nuclear power plant SR structures, for example; SSE, OBE, tornado/hurricane wind, and follow load combinations in AISC N-690. OBE is referred to because, although X-energy uses an OBE equal to one-third of the SSE, section 3.7.3.2.12, Seismic Loads, states that the load combinations with the OBE loads are evaluated for the purpose of comparison. Clarify if modifications to the load combinations or factors in the AISC N-690 load combinations in RG 1.243 were considered in the load combinations in table 3.7-1 for steel NSRST SSCs. X-energy should provide its approach and associated justifications in the PSAR.

Table 3.7-1 (on page 15) has the same number as Table 3.7-1 (on page 14). X-energy should re-number the tables.

B

11 ID Chapter Section Observation Category 28 6

6.4.1 Section 6.4.1.1, Reactor Support Structures (RXSSs) and SR Foundation Functions and Design Criteria, states that Each of the four RXSSs is composed of a steel braced frame that is rectangular in plan. The following paragraph states that The RX support structure is square in plan with approximate centerline dimensions of 8.5 m (28 ft) on each side and 36.6 m (120 ft) tall. If the RX support structure is the same as the RXSS, clarify the shape (i.e., rectangular versus square) of the structure in plan.

Section 6.4.1.1 refers to figure 6.4.1-1, Reactor Support Structure Plan Geometry, and figure 6.4.1-2, Reactor Support Structure Fames - Elevation Dimensions, for the plan and section views of the RXSSs. The dimensions in those two figures are not consistent with the dimensions described in the section. Clarify the inconsistencies and if the figures need to be updated.

Figure 6.4.1-3, Safety Related and Non-Safety Related RB Foundation Configuration, and figure 6.4.1-4, Safety Related And Non-Safety Related RB Foundation Cross-Section, show the combined SR and non-SR foundation configurations. There are several observations pertaining to those two figures:

(a) Section 6.4.1, Reactor Building, states that the SS and the SGSSs are supported by a separate NSRST foundation, separate from that for the RXSSs. That is not clear from the figures and should be clarified.

(b) The dimensions shown in figure 6.4.1-3 are not consistent with the dimensions for the separation between the drilled piers in section 6.4.1.1 for the foundation of the RXSSs. The number of piers in the figure is also not consistent with the number of piers in the text of section 6.4.1.1 and should be clarified.

(c) Figure 6.4.1-4, showing the Reactor Building foundation cross-section, is not clear to understand and should be further explained in section 6.4.1. For example, is figure 6.4.1-4 for the foundation mat, piles and drilled piers for the FHAB and the SS with the basemat of the RXSS shown on the right side of the figure?

B 29 6

6.4.1.1 Section 6.4.1.1 states that the structural systems for the RXSSs braced frames are designed as OCBFs. Confirm if these OCBFs will be designed following the provisions in table 4-1, Acceptable Seismic Force-Resisting Systems for New Nuclear Facilities, of ASCE 43-19, for acceptable seismic force-resistant systems for new nuclear facilities and, specifically, for OCBFs, which would follow the B

12 ID Chapter Section Observation Category provisions in AISC N690 and AISC 341, as applicable. The PSAR should clearly describe the information regarding which specific standard provisions are used to design the OCBFs.

30 6

6.4.1.1 Section 6.4.1.1 lists codes and standards to be used in the design of the RXSSs and SR Foundation.

The list includes ASCE 7-16. The PSAR should describe how ASCE 7-16, which has not been endorsed by the NRC staff, would be used in the design of the SR RXSSs and SR Foundation.

B 31 6

6.4.11 The PFA subsystem of the RCCS is categorized as NSRST SSCs and will be designed and constructed to ASME B31.1 and other industry standard codes.

Per RG 1.87, ASME B31.1 may be used for NSRST SSCs, but X-energy should provide a justification to demonstrate that the code selection is appropriate considering the specific safety significance of the NSRST SSCs, and applicable SSC failure modes for the specific loading condition, material, and environment effect.

B 32 6

6.4.11 PDC 36 is referenced, but the PDC has been subsumed by PDC 6 in the Xe-100 PDC topical report.

(editorial)

C 33 6

6.4.11 The list of components that make up the standpipe functional area in the first bullet of the section contains riser headers twice. The duplicate should be deleted. (editorial)

C 34 6

6.4.11.1 Section 6.4.11.1, Reactor Cavity Cooling System Functions and Design Criteria, describes temperature limits that will ensure that the SR components of the RCCS can perform their RSF. Two questions/perspectives to consider are:

(a) How/when are these temperature limits applicable? Are they applicable in operating/normal/standby conditions (i.e., like a technical specification) or are they performance metrics to be compared to in accident evaluations?

(b) The RCCS is a boil-off system that is vented, so theoretically it should operate near saturation temperature when it is being relied upon to perform its RSF, regardless of how much heat (heat flux) removal it is being relied upon. Would another criterion (i.e., boil-off rate, heat flux, available water inventory, etc.) be more appropriate than (or included in addition to) temperature? How is temperature an appropriate limit for a system that operates at saturated conditions?

B

13 ID Chapter Section Observation Category 35 7

7.3.1.1 Section 7.3.1.1, Shield Structure, states that the SS houses components of the MPS, including the RPVs, SGs, and RCCSs. However, it does not provide a comprehensive list of components that comprise the MPS. X-energy should describe the MPS clearly in the PSAR (e.g., chapter 1, chapter 6, and/or chapter 7) and what specific MPS components are housed in the SS.

C 36 7

7.3.1.1.1 Section 7.3.1.1.1, Shield Structure Foundation Functions and Design Criteria, states that the SS is classified as NSRST for preventing challenges to SR SSCs that perform RSF 1, RSF 1.1, RSF 1.2, and RSF 1.4. Explain why RSF 1.3 (Control Water/Steam Ingress) is not listed. Also, clarify why RSF 1 is listed in addition to RSF 1.1, RSF 1.2, and RSF 1.4. This could be interpreted as including RSF 1.3.

C 37 7

7.3.1.1.1 Section 7.3.1.1.1 addresses functions and design criteria for the SS. However, only some design criteria are described. For example, the applicable PDC(s) corresponding to NSRST PSF 1.4.2.1 needs to be included. In addition, the applicable design criteria should be discussed in more detail in the PSAR. For example, additional design criteria should be included for the use of the multi-wall precast concrete approach for the SS superstructure.

B 38 7

7.3.1.1.1 Section 7.3.1.1.1 uses the phrase increased structural robustness. For example, it states that The increased structural robustness provides a more stable and survivable building during larger seismic events. It appears the phrase is intended to compare with an unknown baseline structural robustness. X-energy should clarify the use and meaning of the phrase and/or revise it.

B 39 7

7.3.1.1.1 Section 7.3.1.1.1 identifies the codes and standards that X-energy plans to use for the design of the SS. It also states that the seismic design of the walls will use the ductility permitted in chapter 12 of ASCE 7-16 and that the structural design of the precast walls will use the provisions for precast concrete walls in chapter 18 of ACI 318-14. The PSAR should justify the use of these code provisions for precast and slabs with the thickness of those in figure 7.3.1-3, Shield Structure Elevation View, which includes walls that are 2.3-meters, 1.4-meters, and 1.0-meter thick as well as floor or roof slabs that are 2.3-meters, 1.2-meters, 1.0-meter and 0.8-meter thick. This should include the performance of the assembled system including connections for walls and slabs with those thicknesses. The PSAR should describe the SS including X-energys understanding of load paths for seismic loading and applicability of codes and standards provisions for precast structures to walls and slab thickness as shown in the PSAR. The PSAR should describe the proposed approach in sufficient detail and clarity for the NRC staffs review, given its novel nature in nuclear power plants.

A

14 ID Chapter Section Observation Category 40 7

7.3.1.1.2 Section 7.3.1.1.2, Shield Structure Performance and Operation, states that the SS is adjacent to the FHAB and refers to figure 7.3.1-1, Project Long Mott Site Layout,. However, the SS is not in the figure; instead, the RB is shown in the figure. The NRC staff understands that the SS encloses the RB, which comprises the four reactor support structures and the four steam generator support structures. X-energy should include additional information regarding the SS (e.g., revised figure and description of the SS in relationship to the RB) in the PSAR.

C 41 7

7.3.1.1.2 Section 7.3.1.1.2 states that the NSRST SSCs in close proximity to a Seismic Category I SSC required for safe shutdown (or to perform required safety functions in the case of the approach used for this project) will adhere to guidance in RG 1.29 by allowing ductile behavior and preclude catastrophic failure. The PSAR should elaborate on how the ductile behavior is to be achieved for the SS when using the multi-wall precast approach.

B 42 7

7.3.10 ASME BPVC-VIII, -XIII, B73.1, B31.1, PTC 19.2, PTC 19.3, and PTC 19.5 are listed in table 7.3.10-2, Reactor Building Cooling Water System Codes and Standards. X-energy should provide more details about the SSCs designed to these codes and the safety classification of the SSCs. X-energy should also provide justifications for the code selections for the construction of the NSRST SSCs to demonstrate that the code is appropriate considering the safety significance of the specific NSRST SSCs and applicable SSC failure modes for the specific loading condition, material, and environment effect.

B 43 7

7.3.10 RCSS is a SR system that is described only in chapter 6 and supports RSF 1.1 and RSF 1.1.2.

Justify why the NICW subsystem RBCW is classified as NSRST, when it is providing cooling to the equipment in the RCSS.

B

15 ID Chapter Section Observation Category 44 8

8.1.4 In a document provided in the eRR, X-energy identified that section 8.1, Quality Assurance, was revised to address the category A observation #117 regarding design D-RAP from the previous readiness assessment.

(a) Section 8.1.4, Design Reliability Assurance Program, states that the D-RAP is implemented by the Quality Assurance Program and the risk-informed performance-based NEI 18-04 processes are leveraged to meet the recommendations in SRM-SECY-95-132 and the subsequent guidance described in SRP 17.4. However, this section does not discuss the needed details about the D-RAP and no documents on D-RAP were made available in the eRR.

(b) The D-RAP is related to pre-operational activities, i.e., design and construction aspects before fuel loading. It should be in place at the CP application stage, and should be described in sufficient detail in the PSAR for the NRC staffs review. The detail will allow the NRC staff to make a finding that the D-RAP in place for the CP application meets the Commission Policy in the SRM. The detailed plant records describing the program should also be available for the NRC staffs audit.

This section on D-RAP has the following:

The purpose of the D-RAP is to provide the processes and programmatic controls for ensuring SSC reliability during the plant design phases. This includes assurances that:

The plant is designed, constructed, and operated in a manner that is consistent with the reliabilities and capabilities need to meet the risk target and defense-in-depth adequacy.

SR and NSRST SSCs do not degrade during plant operations The purpose above is for a reliability assurance program as a whole that includes post-operational aspects and thus goes beyond that of a D-RAP. This section should be revised to be consistent with the purpose of a D-RAP instead of that of a RAP.

A

16 ID Chapter Section Observation Category 45 8

8.5 In a document provided in the eRR, X-energy identified that section 8.5, Environmental Qualification Program, was revised to address category A observation #121 from the previous readiness assessment regarding the environmental qualification of all safety-significant SSCs beyond electrical equipment.

This section states that non-electrical safety-significant SSCs are qualified for their environment based on specific and unique regulatory requirements, codes, and standards.

The NRC staff observes that this general statement, which refers to unidentified regulatory requirements, codes, and standards, is insufficient to meet the 10 CFR 50.34(a) information needs for a CP application. PSAR section 8.5 should identify the regulatory requirements, applicable codes and standards, and provide justifications for how non-electrical safety-significant SSCs will function in the environmental conditions which they are exposed to.

A 46 8

8.5.2.1 The applicant states that the approach for EQ is developed based on guidelines provided in IEC/IEEE 60780-323 and applicable codes and standards referenced in IEC/IEEE 60780-323.

Section 8.5 includes information related to seismic. For the NRC staff to complete a regulatory finding on 10 CFR 50.49, regarding equipment subject to 10 CFR 50.49, other applicable standards and/or regulatory guides should be included in the PSAR. For example, the relevant items for cables could be RG 1.211 and IEEE 383.

B 47 10 Chapter 10 In review of the information contained in section 10.1, Radiation Protection Program, section 10.1.1, Radiation Surveys and Monitoring, and section 10.1.2, Access Control and Dosimetry, the NRC staff requests that the applicant consider addressing the timing of information that will be provided for radiation monitoring details. The NRC staff observes that these sections identify elements of the radiation protection program that will be provided at the operating license stage. However, it does not identify radiation monitoring as one of the elements. X-energy should clarify when this element will be addressed (CP vs. OL application).

C 48 10 Chapter 10 The draft PSAR references ANSI/ANS 6.1.1-2020, Photon and Neutron Fluence-to-Dose Conversion Coefficients. The ANSI standard uses dose conversion factors outside the scope of the 10 CFR Part 20 methodologies. X-energy should clarify how it is proposing to use this standard within their application.

B

17 Table of Acronyms - Project Long Mott Draft PSAR Readiness Assessment Observations Acronym Phrase AOO Anticipated operational occurrence BDBE Beyond design basis event CFR Code of Federal Regulations CP Construction permit DBE Design basis event DBHL Design basis hazard level D-RAP Design reliability assurance program DRS Design response spectra EAB Exclusion area boundary eRR Electronic reading room EQ Environmental qualification ESIM Effective seismic input method F-C Frequency-consequence FEA Finite element analysis FHAB Fuel handling annex building FSAR Final safety analysis report GMRS Ground motion response spectra ISRS In-structure response spectra LBE Licensing basis event LPZ Low population zone LS-DYNA Finite element software for dynamic structural analysis MCER Maximum considered earthquake MPS Main power system NEI Nuclear Energy Institute NGA-EAST Next generation attenuation for central and eastern north America NIAB Nuclear island auxiliary building NICW Nuclear island cooling water NRC U.S. Nuclear Regulatory Commission NSRST Non-safety-related with special treatment

18 NTTF Near-term task force OBE Operating basis earthquake OCBF Ordinary concentrically braced frame OL Operation license PDC Principal design criteria PFA Process functional area PGA Peak ground acceleration PRA Probabilistic risk assessment PSF Probabilistic risk assessment (PRA) safety function RB Reactor building RBCWS Reactor building cooling water system RCCS Reactor cavity cooling system RG Regulatory Guide RIL Research Information Letter RIPB Risk-informed and performance-based RPV Reactor pressure vessel RSF Required safety function RX Reactor RXSS Reactor support structure SAP2000 SAP2000 structural analysis and design software SARRDL Specified acceptable radionuclide release design limits SC Steel plate composite SG Steam generator SGSS Steam generator support structures SPRA Seismic probabilistic risk assessment SR Safety-related SRP Standard Review Plan SS Shield structure SSC Structure, system, and component SSE Safe-shutdown earthquake SSI Soil-structure interaction TEDE Total effective dose equivalent

19 Table of References - Project Long Mott Draft PSAR Readiness Assessment Observations Reference Title ACI 318 American Concrete Institute, "Building Code Requirements for Structural Concrete and Commentary" ACI 349 American Concrete Institute, "Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary" AISC 360-16 American Institute of Steel Construction, "Specification for Structural Steel Buildings" AISC N690 American Institute of Steel Construction, "Specification for Safety-Related Steel Structures for Nuclear Facilities" ANSI/ANS 6.1.1-2020 American National Standards Institute / American Nuclear Society, "Photon and Neutron Fluence-to-Dose Conversion Coefficients" ASCE 41 American Society of Civil Engineers, "Seismic Evaluation and Retrofit of Existing Buildings" ASCE 43-19 American Society of Civil Engineers, "Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities" ASCE 7-16 American Society of Civil Engineers, "Minimum Design Loads and Associated Criteria for Buildings and Other Structures" ASCE/SE 4-16 American Society of Civil Engineers, "Seismic Analysis of Safety-Related Nuclear Structures" ASME B31.1 American Society of Mechanical Engineers, "Power Piping" IEC/IEEE 60780-323 International Electrotechnical Commission / Institute of Electrical and Electronics Engineers, "Nuclear Facilities -

Electrical Equipment Important to Safety - Qualification" IEEE 383 Institute of Electrical and Electronics Engineers, "Standard for Qualifying Electric Cables and Splices for Nuclear Facilities" NEI 18-04 Nuclear Energy Institute, "Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development" NUREG-0800 Nuclear Regulatory Commission, "Standard Review Plan (SRP) for the Safety Analysis Reports for Nuclear Power Plants" RG 1.142 Nuclear Regulatory Commission, Regulatory Guide 1.142, "Safety-Related Concrete Structures for Nuclear Power Plants (Other Than Reactor Vessels and Containments)"

RG 1.211 Nuclear Regulatory Commission, Regulatory Guide 1.211, "Qualification of Safety-Related Cables and Field Splices for Nuclear Power Plants RG 1.243 Nuclear Regulatory Commission, Regulatory Guide 1.243, "Safety-Related Steel Structures and Steel-Plate Composite (SC) Walls for other than Reactor Vessels and Containments RG 1.253 Nuclear Regulatory Commission, Regulatory Guide 1.253, "Guidance for a Technology-Inclusive Content of Application Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors" RG 1.29 Nuclear Regulatory Commission, Regulatory Guide 1.29, "Seismic Design Classification for Nuclear Power Plants"

20 RIL 2021-04 Nuclear Regulatory Commission, Feasibility Study on a Potential Consequence-Based Seismic Design Approach for Nuclear Facilities SRM-SECY-95-132 Nuclear Regulatory Commission, "Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems (RTNSS) in Passive Plant Designs (SECY-94-084)"