ML24037A296
| ML24037A296 | |
| Person / Time | |
|---|---|
| Site: | 99902117 |
| Issue date: | 02/07/2024 |
| From: | Office of Nuclear Reactor Regulation |
| To: | X-Energy |
| References | |
| Download: ML24037A296 (1) | |
Text
X-energy Xe-100 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 perceives to be required to meet the information requirements in Title 10 of the Code of Federal Regulations (10 CFR) 50.34(a) and it was not provided in the draft PSAR (including the table of exemptions).
Category B: Items Requiring Additional Information Item that the NRC staff perceives as needing justification or additional information to support a regulatory finding.
Category C: Other Other observations that should be addressed or considered by X ENERGY, LLC. (X-energy) to support the development of a quality application. If unaddressed, together, they could negatively impact the NRC staff's review of the application, including resources and schedule.
Note: (( )) are used to denote redacted proprietary information ID Chapter Section Observation Cateaorv 1
General General The following items were identified as gaps (X-energy self-identified) for the content within the A
scope:
- Many 'placeholder' items in {yellow highlight} exist that are to be completed by the construction permit (CP) application.
- Some items do not provide any information (e.g., {TBD} or {To be added}).
ID Chapter Section Observation Cateaorv 2
General General There are Commission policies that are applicable to non-light water reactors (non-LWRs).
B Although they are not regulatory requirements, these applicable Commission policies are expected to be addressed by applicants and reviewed by the NRC staff. Some high-level Commission policies such as the Reactor Safety Goal Policy Statement (e.g., Quantitative Health Objectives) are being addressed by X-energy by following the Licensing Modernization Project guidance documents (Regulatory Guide (RG) 1.233 and Nuclear Energy Institute (NEI)18-04a).
However, there are other Commission policies that may be applicable to the plant design for Project Long Mott. An example is the recent Commission policy on digital instrumentation and control common cause failures (SRM-SECY-22-0076). The applicant should explain how it is identifying and addressing these Commission policies.
3 General General The description of certain NST (non-safety-related (SR) with no special treatment) structures, B
systems, and components (SSCs ), classified according to NEI 18-04, appears to be necessary and should be logically organized in the submittal.
The need for such description is partly based on the following:
- To show compliance with applicable regulatory requirements;
- To explain the design bases of the principal design criteria for normal operations (e.g., SSCs that might be described in Chapters 9 and 1 O);
- To provide a high-level plant description in Chapter 1 of the PSAR and support understanding the intended use of the reactor;
- To explain the independence of the SR and/or non-safety-related with special treatment (NSRST) SSCs from NST SSCs;
- To explain the defense-in-depth (DID) Layer 1 NST SSC provisions for overall plant reliability and availability that achieve a frequency of plant transients consistent with owner-operator performance objectives (NEI 21-07, Section 4.2.1.2, "Layers of Defense Evaluation");
- To explain the appropriateness of SSC classification; ID Chapter Section Observation Cateaorv
- To show the design requirements to protect all SR SSCs from any adverse impacts of any design basis hazard levels (DBHLs). This may lead to design requirements to prevent any adverse impacts from the failure of an SSC classified as NST or NSRST that could otherwise prevent an SR SSC from performing its required safety functions (RSFs). (See NEI 21-07, Section 6.1.1 );
- To explain how the NST SSCs are modeled in probabilistic risk assessment (PRA), or non-design basis analyses (non-OBA), and what key assumptions are made to the operation of NST SSCs; and
- To explain how the NST SSCs operates as part of the overall plant operation.
X-energy should clarify how the NST SSC descriptions would be organized in the PSAR and/or auditable plant records.
4 General General Project Long Mott's nuclear steam supply system (NSSS) can be used for electricity and/or B
process heat application (as briefly discussed in Chapter 1 ). X-energy should identify if the process heat application (versus/along with electricity generation) will have any impact on analyses, such as licensing basis event (LBE) and non-LBE analyses? Clarify if it has been analyzed in the CP PRA.
5 General General The draft PSAR (e.g., Section 1.3) uses the term "safety case" instead of the established C
terminology in the current regulatory framework, as indicated in draft RG (DG) 14-04, Revision 1, Section C.2, which states:
"... NEI 21-07, Revision 1, includes use of the terms "affirmative safety case," "safety case," and "licensing case." To avoid confusion and potential unforeseen consequences, applicants using NEI 21 -07, Revision 1, should instead continue to use the established terminology in the current regulatory framework, including use of "safety analysis" and "licensing basis"."
6 General General The draft PSAR refers to several white papers for details or additional information. For example, C
Section 3.1 states that: "Much information related to methodologies and analysis exist in other licensing engagements such as licensing topic reports and white papers. Where appropriate, only ID Chapter Section Observation Cateaorv a high-level summary of such topics is provided without appropriate references to the separate licensing documents."
The NRC staff notes the following two specific examples where X-energy is applying this approach in the draft PSAR:
- Section 3.1 refers to a white paper on PRA technical adequacy multiple times.
- Section 11.3, "Conduct of Operations," refers to a white paper on maintenance staff optimization.
As noted in the NRC staffs feedback documents in response to X-energy's white papers that have been submitted to date, the NRC staff makes no regulatory findings in response to white papers. As such, if there is information in a white paper that is necessary or required in a licensing application, the information should be included (not referenced) as part of the licensing application (e.g., PSAR) for a formal regulatory finding by the NRC staff.
7 General General Many of the diagrams and schematics in draft PSAR are not readable and get blurry when C
zoomed in. For an example, see Figure 7.3.8-1, "HCS Schematic."
8 General General Section 6.4.6, "Reactor Pressure Vessel," states that the reactor pressure vessel (RPV) is B
designed considering vibrations and flow-induced vibrations (FIV) occurring at full-power and during normal operating transients. FIV considerations are not discussed for other SR and NSRST components. Per RG 1.20, Revision. 4, "Comprehensive Vibration Assessment Program for Reactor Internals during Preoperational and Startup Testing," FIV, acoustic resonance (AR),
acoustic-induced vibration (AIV), and mechanical-induced vibration (MIV) could cause failure of the reactor internals and even steam generator components.
The NRC staff expects that an appropriate assessment of vibrations including FIV, AR, AIV, and MIV will be performed for the structural integrity analysis of relevant SR and NSRST components including the RPV.
The NRC staff also suggests that a discussion of such vibration analysis methods be ID Chapter Section Observation Cateaorv appropriately added to Section 3.1.4, "Other Methodologies and Analyses."
In addition, per RG 1.20, vibration measurements should be performed during the pre-operational tests, including testing before fuel loading and start-up testing, to confirm the predictive vibration analysis and an inspection program should be developed to address both quantitative and qualitative verification of the predictive analysis and measurement program results. The NRC staff suggests that the vibration measurement and inspection plans for the CVAP program be discussed in Chapter 8 and Chapter 12.
9 General All Probabilistic risk assessment (PRA) safety functions (PSFs) are discussed in multiple Chapters.
C applicable For example, several tables in Chapters 5, 6, and 7 include specific PSFs (e.g., PSF 1.2.2).
Section 5.2, "Required Safety Functions," provides an overview of individual RSFs consistent with NEI 21-07. However, there is no section or a location that provides a similar discussion for the PSFs. Although the PSFs are part of the PRA, an overview discussion, such as the overall structure and the numbering scheme, of the PSFs should be provided in one location. For example, when PSFs are mentioned in Chapters 5, 6, and 7, this overview discussion should support the NRC staffs overall understanding of the PSFs. One logical place may be in Chapter 5 subsequent to the discussion on RSFs in Section 5.2.
10 1
1.1.2 Regarding the following statement in DG-1404, Revision 1, C.2 Staff Position d.:
C "In addition to the information identified in NEI 21-07, Revision 1, Section C.1.1.2, on intended use of the reactor, applicants should also provide the nature (e.g., physical form) and inventory of contained radioactive materials."
The nature and inventory of contained radioactive materials should be added to Section 1.1.2, "Intended Use of the Xe-100."
11 1
1.1.3 Section 1.1.3, "Overall Configuration," includes the following statement: "{If the plant includes C
more than one reactor, the relationship of the reactors should be described, including major dependencies such as shared systems and structures.}"
This is a direct quote from NEI 21-07. Since Xe-100 has four modules (i.e., more than one ID Chapter Section Observation Cateaorv reactor), the PSAR should include the subject discussion.
12 1
1.1.4 This section only discusses Section 1.1.4.1, "Reactor Systems and Components." This section A
should also discuss the remaining plant SSCs for the Nuclear Island, the Conventional Island, and some aspects (e.g., interfaces with the Conventional Island) of Seadrift Operations, to facilitate an overall understanding of the plant.
In NEI 21-07, the following are listed as examples to be discussed in addition to Section 1.1.4.1:
Section 1.1.4.2, "Secondary Systems and Components"
- Heat transfer and cooling system
- Power conversion system
- Power transmission (e.g., switchyard)
Section 1.1.4.3, "Significant Support Systems and Components"
- Fuel handling
- Fuel management, including spent fuel storage
- Control room
- Electrical power
- Radioactive waste Section 1.1.4.4, "Major Structures"
- Reactor building
- Auxiliary, secondary, and support buildings
- Cooling towers/systems
- Co-located facilities (e.g., cogeneration, fuel processing, and buildings)
The discussion in Section 1.1.4, is preliminary but should be "sufficiently clear for the reader to understand the initial plant functionality [NEI 21-07]."
ID Chapter Section Observation Cateaorv 13 1
1.3.1 In addition to the description that discusses NEI 21-07 and RG 1.233 for the methodology, C
X-energy submitted a licensing topical report (L TR), which was subsequently approved by the NRC staff, that contains additional implementation descriptions while remaining consistent with the principles and methodology in NEI 18-04 and RG 1.233.
X-energy should consider discussing this L TR in this section if Project Long Mott plans to reference it as part of its CP application.
14 1
1.3.2 NEI 21 -07 states that: 'The section should begin by establishing the overall performance C
objectives - the regulatory dose criteria and quantitative health objectives (ref. NEI 18-04, Figure 3-1 )." The draft PSAR does not discuss this information and should include it at a summary level.
15 1
1.3.3 Section 1.3.3 of NEI 21-07 states:
A "DID is a key element of the LMP-based affirmative safety case and the demonstration of reasonable assurance of adequate protection of public health and safety. In this overview, the applicant should summarize the overall conclusions for each of the three DID elements: plant capability, programmatic capability, and integrated RIPB DID adequacy evaluation results, which together establish the DID baseline described in detail in Chapter 4."
The draft PSAR does not provide a summary of the conclusions of the plant capability DID evaluation, which is within the scope of the CP application. Although it is preliminary and PSAR Section 4.2, "Defense-In-Depth," is mentioned for the details, X-energy should consider providing the summary in this section.
16 1
1.4.1 DG-1404 states that: "In Chapter 1 of the SAR, in addition to the information identified in A
NEI 21 -07, Revision 1, Section C.1, applicants should include summary tables with the following information, which appears in full elsewhere in the SAR:
(1) The generic safety issues, unresolved safety issues, and Three Mile Island (TMI) action items technically applicable to the design, and their proposed resolution (for generic safety issues, see NUREG-0933, "Resolution of Generic Safety Issues" (Reference 21 ). The guidance on ID Chapter Section Observation Cateaorv applicability of regulations in Appendix B to the Advanced Reactor Content of Application Project (ARCAP) Roadmap interim staff guidance (ISG) may provide useful insights in this area."
The draft PSAR does not provide the summary table. This summary table would support the efficiency of the NRC staff's review by centrally locating this information. A summary, or a pointer to other sections that discusses them in detail, should be included in the table for those that are found to be technically applicable.
17 1
1.4.4 Table 1.4.4-1, "Licensing Topical Reports and Additional References Incorporated by Reference,"
C discussed in this section, is missing.
18 3
All The event sequences and specific occurrences/failures during events should be described in B
applicable more detail. For example: What is a "pump down of primary system"? (Are helium circulators still running during the leak?)
Time units on plots/figures should be converted to more meaningful units (i.e., using hours or days rather than hundreds of thousands of seconds).
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All The break size assumption should be justified (shown to be bounding). Describe how the B
applicable appropriate break sizes to be analyzed were determined (maybe reference to a TR/methodology?).
This observation applies to small depressurization (SD), steam generator tube leak (SGTL), and any other LBEs that involve assumed "break" or leak sizes.
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3.1.2 Instead of describing the mechanistic source term (MST) methodology per NEI 21 -07 B
Section 2.2, it refers to the X-energy MST L TR. It is noted that the MST L TR, Revision 1 was shared with the NRC staff in April 2021. X-energy plans to submit the MST L TR for the NRC staff's approval in January 2024.
X-energy should include additional information on how it intends to meet NEI 21-07, Section 2.2.
In addition, the PSAR should include a discussion of how X-energy is leveraging the MST L TR to address the information pertaining to this section.
ID Chapter Section Observation Cateaorv 21 3
3.1.2 DG-1404, Revision 1, Section C.3.c, specifies two options to demonstrate that the facility meets A
10 CFR 50.34(a)(1 )(ii)(D). The PSAR does not state which of these two options is being used and whether an exemption is needed. It is not clear how X-energy intends to meet DG-1404 Section C.3.c.
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3.1.3 The information included in Section 3.1.3 appears as more of a summary in nature rather than B
providing indications of what methods are being used for specific purposes in the OBA analyses.
Specific codes and their uses should be identified, similar to Section 3.1.2 (references to separate reports, as in Section 3.1.2, seem appropriate).
Further, the referenced transient safety analysis methodology TR, which was approved with limitations and conditions, is only X-energy's plan to implement the methodology; it did not approve implementation or application of specific codes. That said, the information in this section appears accurate, but additional details are needed to understand the OBA methods used as part of the application.
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3.1.4.1.2.5 The draft PSAR specifies the normal thermal load and it states that it is not applicable to the B
reactor buildii (RB). The RB internal con.
structures normal operating temperature is defined as ((
11 degrei Celsius (°C) (((
11 degrees Fahrenheit (°F)). (1) Clarify how Xe-100 is maintaining ((
11°C at all locations for all RB structures for normal operating conditions. (2) Clarify how temperature is maintained within code limits at all locations for all RB structures during accident conditions. A review of Chapter 6 did not provide this information.
How does Xe-100 plan to address the temperature for normal operating and design basis accident loading conditions, global and local temperature distribution, gradient short-term and long-term on structure which are needed for design of structures and components and assessing degradation due to long-term operation?
Also, how does Xe-100 plan to address the end-of-life (EOL) bounding accumulated Neutron Fluence, Gamma Dose, and Temperature on structures which are needed for assessing and evaluating degradation due to long-term operation?
ID Chapter Section Observation Cateaorv 24 3
3.1.4.4 There is some inconsistency in the discussion regarding the components to be included for C
stress analysis:
- The first paragraph states that Section 3.1.4.4 presents the methods of analysis for metallic components that perform the maintain core geometry RSF and Table 3.1.4.4-1, "Pressure Boundary Component Classes," lists these components. It is noted that the components listed in Table 3.1.4.4-1 include components besides those required to maintain core geometry (e.g., the secondary side components). X-energy should consider removing from the scope of this table components (or portions thereof) that do not perform the listed RSF.
- Section 3.1.4.4, "Stress Analysis" (including Piping Analysis)," is a summary discussion of stress analysis and covers the various components that are discussed in detail in subsequent subsections. However, stress analysis methods for the reactor cavity cooling system (Section 3.1.4.4.9, "Stress Analysis of the Reactor Cavity Cooling System,") and for NSRST SSCs (Section 3.1.4.4.10, "Stress Analysis of Non-Safety Related with Special Treatment SSCs,") are not discussed. For consistency, X-energy should consider addressing these components in Section 3.1.4.4.
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3.1.4.4 The mechanical performance of the TRI SO-fuel particles and fuel pebbles needs to be evaluated C
by testing and addressed in the fuel qualification program. Therefore, it is suggested that this information be mentioned in Section 3.1.4.4, to inform the reader that the mechanical performance of all the SR and NSRST components will be evaluated to ensure the structural integrity of these components.
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3.1.4.4.7 X-energy should clarify if additional data are needed to justify the extrapolation of stress B
allowables beyond 300,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
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3.1.4.4.7 Clarify if X-energy is planning to apply Subsubarticle HBA-2610 to components other than "small B
products" as defined by HBA-2610.
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3.1.4.4.7 One of the technical areas identified in NEI 19-03, "Advanced Reactor Codes and Standards B
Needs Assessment," is elevated temperature.
ID Chapter Section Observation Cateaorv X-energy should consider addressing the impact of elevated temperatures on safety-significant structures in this section.
29 3
3.1.4.4.8 What references support the graphite stress analysis method in Section 3.1.4.4.8, "Stress B
Analysis of Graphite"? More details on the method should be provided to support the NRC staff's evaluation of the method.
30 3
3.1.4.4.9 This section states that American Society of Mechanical Engineers (ASME) AG-1 2019 edition B
will be used for the reactor cavity cooling system (RCCS) stress analysis.
ASME AG-1 Code is the Code on Nuclear Air and Gas Treatment. X-energy should provide additional explanation and justification for using this code for reactor cavity cooling system stress analysis.
31 3
3.1.5 This section states: "There is robust operational history of HTGRs including Peach Bottom Unit 1 B
and Fort St. Vrain in the USA. Also, Arbeitsgemeibschaft Versuchsreacktor (AVR) and Thorium Reactor (THTR) in Germany and Hogh Temperature Reactor (HTR) HTR-10 in China. ANL and Texas A&M Univ provided data on RCCS performance for air cooled systems similar to the Xe-100 RCCS. {Will point to data feed into RG 1.203}."
Is there information available from the operating history of these reactors pertaining to temperature, pressure, and radiation impact on structures? X-energy should consider addressing in the PSAR.
32 3
3.2 SD-3, an anticipated operational occurrence (AOO) shown in Figure 3.2-1, "Comparison of LBEs C
against the Frequency-Consequence Target," does not show up in Table 3.2-1, "Summary Table of AOOs, BDEs, and BDBEs." However, SD-3 shows up in Table 3.2-2, "30-day EAB TEDE for AOOs, BDEs, and BDBEs." Further, the number of LBEs listed in Table 3.2-1 and Table 3.2-2 is different although it should be the same for both tables. X-energy should reconcile these two tables in the PSAR.
In addition, SD-3 is a risk-significant LBE according to Figure 3.2-1, "Comparison of LBEs to the Frequency-Consequence Target." This is currently not consistent with the statement of 'no risk-ID Chapter Section Observation Cateaorv significant LBE' discussed on Page 3.2-8. X-energy should reconcile the discrepancy in the PSAR.
33 3
3.2 In the licensing modernization project (LMP) framework (i.e., NEI 18-04 and NEI 21-07), an AOO, C
design basis event (DBE), or beyond design basis event (BOBE) is regarded as risk-significant if the combination of the upper bound (95th percentile) estimates of the frequency and consequence of the LBE are within 1 percent of the F-C Target AND the upper bound 30-day TEDE dose at the EAB exceeds 2.5 mrem.
It appears that the frequency and consequence estimates of the LB Es in Figure 3.2-1 are the mean estimates for the frequencies and the 30-day exclusion area boundary (EAB) total effective dose equivalent (TEDE) consequence values in Table 3.2-2.
The NRC staff observes that, for the purpose of presenting risk-significant LBEs, the upper bound (95th percentile) estimates of the frequency and consequence of the LB Es should be placed onto Figure 3.2-1 and compared with the cross-hatched region (within 1 percent of the F-C Target). For example, the upper bound estimates of SD-3, an AOO, is expected to be closer to the F-C Target, making this LBE to be more risk-significant (i.e., closer to the F-C Target).
(The NRC staff observes that the upper bound estimates are yet to be developed (i.e., {TBD}) in Table 3.2.1.1 -2, "30-Day EAB TEDE for AOOs, DBEs, and BDBEs.")
34 3
3.3 Information provided for the event sequences available at this stage of the readiness review C
seems reasonable. Events may require additional context beyond "below the reporting threshold" in the figure of merit section (e.g., if there is no stated release, what was the figure of merit that yielded a conclusion of no release?).
35 3
3.3.1.1, TT-001 and TT-002 have different turbine bypass valve (TBV) opening rates and delays. Is this B
3.3.1.2 by design (i.e., influenced by the control system), or some other assumption associated with the events? Those inputs should be explained.
36 3
3.3.4.1.1 The loss of offsite power (LOOP) event description (Section 3.3.4.1.1, "Loss of Offsite Power with B
Turbine Generator (TG) Maintaining House Load,") states that emergency power generators would have to be relied upon to support decay heat removal (if the main generator is unavailable ID Chapter Section Observation Cateaorv to support house loads). This should be clarified. For which LBEs does X-energy intend to use emergency power generators? How does this align with the standby power source(s) described in Chapter 7, and the safety classification of those generators?
37 3
3.3.5 CRW (control rod withdrawal) events (AOOs and other LBEs) start with a rod withdrawal and B
reactor power increase, but the fuel temperature response described says that the fuel temperature decreases from the onset of the event. The event sequence predicts ~2-3 secs between the beginning of rod withdrawal and the high flux trip. X-energy should clarify if the fuel temperature trends down during this ~3 secs span (prior to the high flux trip) and provide an explanation for this behavior in the PSAR.
38 3
3.3.6.2 LBE SD-004 is an AOO that results in maximum fuel temperatures in excess of ((
))°C, B
which likely raises the probability/rate of fuel particle failures.
- The event sequence (Table 3.3.6.2-1, "SD-004 Sequence of Events,") implies that it takes
~8 mins to reach the reactor trip setpoint. Has consideration been given to whether the consequences (fuel temp, etc.) of such an event could be meaningfully mitigated by earlier detection and shutdown?
- Figure 3.3.6.2-13, "SD LBE Short-Term Reactor Power vs Fuel/Moderator Average Temperatures," is supposed to be reactor power vs. Fuel/Mod temperature, but the actual figure provided seems to be a copy of Figure 3.3.6.2-11, "SD LBE Long-Term Reactor Power."
- Figure 3.3.6.2-15, "SD LBE Long Term Reactor Core Temperature," seems to depict a higher maximum fuel temperature than what is summarized in Table 3.3.6.2-2, "SD-004 Sequence of Events."
The text describing the trends shown in Figure 3.3.6.2-16, "SD LBE Long-Term RPV Temperatures," does not address the phenomena seen related to reactor outlet temperature (ROT) after ~ 700 secs. X-energy should clarify what is happening at that time to drive the depicted temperature fluctuations.
39 3
3.4, 3.4.15 PSAR Section 3.4, "Design Basis Events," states that: "due to the complex nature of the models, B
ID Chapter Section Observation Cateaorv the evaluations are computationally limited to being used only for the first few hours of these events." This "plant transient phase" terminates in some cases where the transient is not stable or decreasing (e.g., SD-004), and then passes off to a "long-term plant response" phase, described at a high level in PSAR Section 3.4.15, "Long-Term Plant Response to AOOs, DBEs, and BDBEs," which is not complete at this stage.
To determine a transient is over from a consequence perspective, figures of merit should be in a stable condition, with none degrading or undergoing transitory behavior. In cases where a transient spans two models, a justification should be provided for any handoff in calculation of consequence. Lacking additional information at this stage, the NRC staff expects this subject to be an area of focus during the CP application review.
40 3
3.4.7.3, For events that combine a helium pressure boundary (HPB) leak that is not isolated (implying the B
3.4.7.4, Helium coolant leaks out) with a form of forced cooling (which relies on helium), the PSAR should 3.4.8 explain how (and to what level/significance/flow rate etc.) forced cooling is maintained.
41 3
3.5.4.2 The section title lists a circulator runback, but the event description states that it involves a C
circulator trip. This discrepancy should be clarified.
42 3
3.8 The draft PSAR states that the stress analyses will be documented in the ASME Design Reports B
for each component in the final safety analysis report (FSAR) and will meet Section Ill, Division 5 of the 2023 edition of the ASME Code with deviations from RG 1.87, Revision 2 justified.
The NRC has not endorsed the 2023 edition of Division 5. The PSAR should include a discussion on the deviations and justifications.
43 3
3.8 X-energy should clarify if helium bypass flow is considered in the stress analyses of the core B
support structures and/or graphite.
X-energy should clarify if the impact of the expansion and contraction of graphite has been considered on the core metallic structures.
44 3
3.10 Section 3.1 0.1.4, "Radiological Consequence Analysis," states that: "The method used to B
ID Chapter Section Observation Cateaorv quantify the consequences is the same as the method used for the DBAs in Section 3.6... "
The statement is not clear to the NRC staff. In which ways are the methods the same? Additional details should be added regarding how the consequences are quantified that support the results.
The NRC staff notes that Section 3.6, "Design Basis Accidents," does not describe a methodology for quantifying consequences. Instead, it refers back to PSAR Section 3.1,
"Methods and Analysis," and Section 3.1.1, "Probabilistic Risk Assessment."
X-energy should clarify in the PSAR if non-core related source radiological analyses will use the methodologies described in Section 3.1.2, "Mechanistic Source Term Methodology" and/or Section 3.1.3, "Design Basis Accident (OBA) Analytical Methods."
45 3
3.10 Non-Core Related Source Radiological Analysis B
Fuel and non-Fuel related sources:
X-energy should clarify if it has considered EOL bounding accumulated Neutron Fluence, Gamma Dose, Temperature for structural performance in this area.
46 3
3.10.4.1.2 Thermal Event for Spent Fuel in the Storage Phase:
B "An evaluation of the Xe-100 spent fuel interim storage (SFISF) demonstrates that the fuel temperature response to a ventilation blockage event for a period of 7 days is less than 300°C.
There is sufficient margin for fuel failure. The analysis results shows that fuel temperatures peak around four days as heat from the fuel is transferred to the surrounding building."
What is the allowable temperature on structural SSCs? It appears 300°C is considered high from a structural performance standpoint. X-energy should clarify if the fuel peak temperature presents a challenge to surrounding structures.
47 3
3.12 Aircraft Impact Assessment - The last paragraph of this section has the following:
C
"{Add details to the section on how the design reduces the impact of the aircraft impact including ID Chapter Section Observation Cateaorv embedded structures, etc. This will be part of the initial calculation assessment}. Additional details will be provided with the FSAR."
The NRC staff expects that the initial calculation will be available for audit during CP application review.
48 3
3.14 Regarding the following references:
B 11. Xe-100 Civil/Structural Design Criteria report. The NRC staff expects that the design criteria report will be available for audit during CP application review.
- 12. American Society of Civil Engineers (ASCE) Standard 43-05 is referenced in the document.
X-energy should consider using the later version of the standard (ASCE 43-19) to capture the latest information in this area.
49 3
Table of The subsections under 3.1.4, "Other Methodologies and Analyses," list various 'other C
Contents methodologies and analyses' (e.g., Section 3.1.4.1, "Civil and Structure Analysis Methodologies,"
and Section 3.1.4.3, "Electrical Load Analysis," etc.). The NRC staff suggests that the Chapter 3 Table of Contents include these subsections, which should facilitate the NRC staff and other stakeholders to easily locate the specific methodologies and analyses up front in the chapter.
50 3
Figure Helium circulators trip off near the beginning of the event, but coast down to ~500 rpm, not B
3.4.12-5 0 rpm. X-energy should clarify in the PSAR what keeps the circulators turning at 500 rpm?
51 3
Table 3.3.1 Decay heat is specified as "curve fit to DIN 25485 standard results." A justification for the B
applicability of this to the X-Energy design should be provided.
52 3
Table The TT-002 sequence of events table is mis-named as TT-001, and it does not include any C
3.3.1.2-1 information regarding core temperatures during the event sequence.
53 3
Table 3.3-1 Table 3.3-1 contains references to Note Numbers (i.e., Note 1 and Note 6), but the notes C
themselves are not included in/near the table.
ID Chapter Section Observation Cateaorv 54 3
Table 3.3-1 For which LBEs are bounding/conservative initial conditions assumed? (i.e., Assume ~102 B
percent reactor thermal power (RTP), etc.). Is 199.6 MWth appropriate for core thermal power, given the reactor is described as a 200 MWth unit? Similarly, does predicted control rod behavior (trips) consider having the highest worth rod stuck out? Do these requirements change between AOOs and DBEs/BDBEs or DBAs? X-energy should consider adding these details in the PSAR.
55 4
Table 4.2-8 Table 4.2-8, "Plant Capability DID Evaluation for Prevention and Mitigation," Section 4.2.1.3, B
"Single Feature Reliance," and Section 4.2.1.4, "Prevention-Mitigation Balance."
Overall, the discussions and the content or format can be enhanced. For example:
- The title of the table should contain something about single feature reliance.
- The table may benefit from categorizing the LBEs. For example, Section 4.2.1.3, "Single Feature Reliance," discusses four DBEs with single feature reliance: LOPF 1, LOPF 2, LOPF 6, and LD-1. It may be highlighted in a category. The rest of the table appears to cover risk-significant LBEs for which special treatments were added to provide assurance against over reliance on any single feature across multiple layers of defense and the DID objectives requiring those actions.
The statement above the table is followed by: "No LBEs meet the criteria for risk significance per PSAR Section 3.2." It appears that the table can change significantly if this bracketed information is confirmed later. This expression is used multiple places in Chapters 3 and 4, which makes the NRC staff's assessment challenging.
56 4
Tables 4.2-Tables 4.2-5 through 4.2-7 should include further explanation regarding what information they are B
5, 4.2-6, intended to convey. It is recognized that the actual table values/entries may change (i.e., they are 4.2-7 yellow-highlighted), but the general message/format of the tables is unclear. What are the tables demonstrating regarding DID? Is it the intent to show that DID exists or doesn't exist for specific LBEs?
Table 4.2-5, "Evaluation of AOOs Against Plant Capability Guidelines":
ID Chapter Section Observation Cateaorv
- The term "challenged" should be clarified in this context. X-energy should also clarify how it is decided whether a SR SSC is considered to have been challenged during an event.
- Are the SSCs listed in the RSF columns additional available SSCs that ensure DID exists (i.e.,
they back up the SR SSCs), or do they include all SSCs available to perform that RSF (and assume those not listed are unavailable/failed/Not applicable)? Is this consistent between the different RSF columns? For example, the column "Retention of radionuclides" includes Fuel (SR),
HP8 (NSRST), and NST SSCs, which seems to be comprehensive, but the "Core Heat Removal" column never lists RCCS (SR), which presumably would be available/in service.
Table 4.2-6, "Evaluation of D8Es Against Plant Capability DID Guidelines":
- Same questions as above. Additionally, explain the column that states: "any SR failures to perform their RSF?". Is this effectively asking "are any SR failures required to make this L8E happen?"
Table 4.2-7, "Evaluation of 8D8Es Against Plant Capability DID Guidelines":
- Same question as Table 4.2-6 regarding the rightmost column.
- What does it mean when there is a failure in the RSF SSC column? Does that mean the specific L8E evaluated involved the failure of an RSF? Does it mean that, for that L8E scenario, no DID was available and therefore a single failure could lead to the RSF not being accomplished? Does it mean the SR equipment failed in that scenario?
X-energy should consider these questions and update the PSAR accordingly.
57 5
5.5.1 Section 5.5.1, "Non-Safety Related SSCs Performing Risk-Significant Functions," states "this C
safety classification is based on applying steps 48 and 58 in Figure 5.1-1 and.... " However, Figure 5.1-1 is not included anywhere in PSAR Chapter 5, NEI 18-04, or NEI 21 -07. Is Figure 4-1 from NEI 18-04, Revision 1 meant to be referenced in Section 5.5.1?
58 5
Table 5.2-1 Why does RSF 1.4.1 effectively include two functions? "Maintain HP8 and core geometry" both 8
ID Chapter Section Observation Cateaorv seem significant enough to warrant their own individual RSFs. Many SSCs would likely overlap, but many SSCs associated with the HPB won't necessarily be related to core geometry.
X-energy should consider reformulating this RSF into two separate functions or justify keeping them together.
59 5
Table 5.4-1 The wording of RSF 1 is somewhat "definite" regarding radionuclide retention, when realistically a C
certain amount of release is expected (to be within specified acceptable system radionuclide release design limit (SARRDL)). In practice, some radionuclides (a small failure fraction and/or manufacturing defect fraction) will not be retained during operation, AOOs, DBEs, etc. In order for the RSF to be considered met, some criteria must be met (i.e., the SARRDL). Should this be referenced in the RSF itself, or somewhere nearby in the PSAR explaining that "radionuclide retention" is not "absolute?".
Note: The third paragraph of Section 5.2 does mention specific fuel particle and pebble design criteria (based on the fuel qualification) that helps ensure fuel does not exceed acceptable limits.
The statement is somewhat vague, though.
60 5
Table 5.4-1 Table 5.4-1, "Table of SR SSCs and the LBEs Prevented and Mitigated by RSFs," lists the B
graphite-related RSFs as (1.1, control reactivity) and (1.2, control heat removal). However, the design seems to have the graphite effectively form the inner-most dimensions of the core volume that ultimately forms the overall shape of the pebble bed. RSF 1.4 (maintain core geometry) is credited to metallic structures for the core and the pebble handling system inleUoutlet. Should the graphite be credited for contributing to RSF 1.4 as well?
61 5
Tables 5.4-Some of the features listed in Table 5.4-1 and Table 5.5-1, "Table of SSCs Classified as NSRST C
1, 5.5-1 for DID," are not described and/or differentiated (it is recognized they are highlighted in yellow).
Therefore, it leads to instances where it is unclear whether an SSC is SR or NSRST. For example, what is the difference between "Core Graphite Structures" (Table 5.4-1, SR) and "core structural graphite" or "reflector graphite" (both in Table 5.5-1, NSRST)? X-energy should use consistent terminology for similar components or clearly differentiate the names of the SSCs to facilitate NRC staff understanding of the relation between the safety-significant SSCs and the functions it supports.
ID Chapter Section Observation Cateaorv Supporting note: This question points to the fact that the tables address RSFs and "non-RSF" PSFs. Some systems in Table 5.4-1 are SR, but also perform "non-RSF" PSFs and are therefore also listed in Table 5.5-1. However, the Table 5.5-1 column for the SSCs says "NSRST SSCs,"
even though some of them may be SR based on being relied upon to support a RSF in Table 5.4-1. This may be leading to confusion when looking at Table 5.5-1 and thinking "why is the core structural graphite NSRST?" The NRC staff presumes that, in reality, it's an SR SSC that performs a PSF that wasn't deemed important enough to be included in the "RSF" subset.
62 6
6.4 The PSAR does not include analyses for degradation mechanisms, e.g., stress relaxation A
cracking, irradiation embrittlement, etc. for the metallic structures.
Degradation mechanisms should be analyzed as part of the CP application, e.g., minimizing copper content in ferritic steels, weld constrained considerations for stress relaxation cracking, etc. Section 6.4.3.3, "Core Metallic Structures' Qualification, Testing, and Programs," states that:
'The Environmental Qualification Program is discussed Section 8.5." Section 8.5, "Environmental Qualification Program," does not discuss the environmental qualification program for the core metallic structure. The PSAR should include the information either in Section 8.5 or in Section 6.4.3.3. A testing program might be necessary to help address some of the degradation mechanisms such as stress relaxation cracking.
63 6
6.3 Table 6.3-1, "SR SSC special treatment requirements," does not contain the needed information A
and is to be filled out later at the level of detail allowable by the current design {yellow bracketed}.
In addition, the NRC staff suggests that the codes and standards in this table be consistent with, or evaluated against, those discussed in individual SSC description sections under Section 6.4, "Descriptions for Safety Related Structures, Systems and Components."
64 6
6.4 What edition of Division 5 of the ASME B&PV Code, Section Ill is being used to qualify the B
graphite and graphite carbon fiber composite materials? Are graphite carbon composite materials intended for use in the X-energy reactor?
As stated in the draft PSAR, "Graphite reflector structures are made from multiple forms of graphite and carbon fiber composite materials. All materials conform to the minimum requirements of mandatory Appendix HHA-1 of Subsection HH of Division 5 of the ASME B&PV ID Chapter Section Observation Cateaorv Code, Section Ill."
65 6
6.4.1 Reactor Building (RB)
B Section 6.4.1.1, "Reactor Building Functions and Design Criteria," states:
'The RB structurally protects the geometry for (1) passive removal of residual heat from reactor core to the ultimate heat sink and (2) permit insertion of sufficient neutron absorbers and inherent reactivity feedback to provide safe shutdown during design basis events (DBEs) and design basis accident (DBAs). The RB provides structural support and protection to the Reactor System (RS), Reactor Cavity Cooling System (RCCS), Steam Generators (SGS), Steam Generator Dump System (SGDS), Fuel Handling System (FHS), Reactor Protection System (RPS)."
Figure 6.4.1 -1, "Reactor Building Plan View," Figure 6.4.1-2, "Reactor Building Plan View A-A,"
and Figure 6.4.1-3, "Reactor Building Plan View B-B" (Pages 20-22), should contain additional details. For example: (1) there is no indication of the below grade and above grade elevations, (2) there is no annotation to relate the description and to identify the internal SSCs, (3) the general arrangement and major supports in plan and elevation views are not shown, (4) the extent of reinforced concrete (RC) and steel-concrete composite (SC) structures is not delineated, and (5) it is not clear which part is the reactor citadel.
Provide figures to represent a high-level overview of structural arrangements for the NRC staff to review the CP application.
66 6
6.4.1.1 Multiple statements regarding principal design criteria (PDC) 2 and 4 are made (using slightly C
different language/explanations) near the beginning of Section 6.4.1.1. X-energy should consolidate and/or clarify.
67 6
6.4.1.1 Near the bottom of Page 6.4-23, multiple external hazard design requirements are referenced.
C Almost all include a reference to a regulation or guidance/standard. The statement associated with flood water heights does not include a similar reference, but probably should.
68 6
6.4.1.1 The second to last paragraph on Page 6.4-23 states: "The site design ensures that a ((I)) psid C
ID Chapter Section Observation Cateaorv or,reater overpressure does not cause damage... " Wouldn't a higher overpi ssure (greater than
((
11 psid) bl more limiting/less conservative than a lower one (less than ((
11 psid)? If so, the statement "((
11 psid or greater" is ambiguous and shouj be clarified. Additionally, a reference/justification for the overpressure value used (((
11 psid) should be provided.
69 6
6.4.1.1, It would be helpful if the major features that the RB supports (listed at the end of page 6.4-19, C
Figures into the next page) were located/identified in the figures. The room identifiers in the section views 6.4.1 -1, are not useful by themselves. Consider including a short label with the rooms' purposes and/or 6.4.1 -2, what major SSCs the rooms support. This will help support the bullets in NEI 21-07, Revision 1, 6.4.1 -3 Section 6.4.1 (specifically regarding the SSC location), and also help interpret Figure 6.4.1-4, "Reactor Building Relief Pathway."
70 6
6.4.3.1 X-energy should clarify in the PSAR the bounding EOL fluences for the reactor core barrel (CB)
B structures at temperatures of interest, e.g., minimum temperature, maximum temperature, and temperature at the maximum end-of-life fluence. X-energy should also clarify if the CB structures will be subject to an irradiation surveillance program?
71 6
6.4.3.3 Clarify if the thermal fatigue cycles caused by the demonstration tests will be considered in the B
total number of fatigue cycles anticipated over the reactor's lifetime.
72 6
6.4.4 The absence of a qualification program for irradiated graphite in the PSAR is a gap.
A Slide 6, "Approach to Provide Graphite-Related Safety Case," from the August 24, 2023, Graphite Engagement Meeting (ML23222A270) presented to the NRC staff, lists reports and preliminary analyses which are not identified in the draft PSAR. These reports and/or their contents should be included in the PSAR or submitted for the NRC's staffs review of the CP application.
73 6
6.4.4.1 The NRC staff notes that neither ASME Code Case N-903 nor the 2021 edition of the ASME B
Boiler and Pressure Vessel Code have been endorsed by the NRC. X-energy should provide the justification for its use in the Xe-100 design.
74 6
6.4.4.2 Clarify how helium bypass flow is considered. The draft PSAR states that the outer boundaries of B
ID Chapter Section Observation Cateaorv components containing hot gas are always within a cold helium environment at a higher pressure.
How is thermal fatigue caused by the potential mixing of hot and cold helium if helium bypass flow occurs?
Clarify if the cold helium flow resides outside of the core barrel but inside of the RPV or otherwise.
75 6
6.4.4.2 The draft PSAR includes the following bracketed statement, "Some graphite core components B
may be designed to SRC-11, the safety basis here is under development and will be discussed in the ongoing graphite engagements."
Are there any additional insights that X-energy can provide to the NRC staff on SRC-11 designated graphite? For example, what potential graphite core components are being considered? X-energy should consider the questions and update the PSAR accordingly.
76 6
6.4.4.1 Describe the "labyrinthine style seals" and how they seal the hot gas duct interface in the PSAR?
B What are the materials of construction of the seals? What is the potential for hot helium bypass flow around the seals? X-energy should consider the questions and update the PSAR accordingly.
77 6
6.4.5.1 What material is the ijH constructed of - Alloy ((
)) (a code-approved material under C
Division 5) or Alloy ((
))?
78 6
6.4.5.1 What edition of Division 5 is the Defuel Chute (FUCH) designed and constructed to? The draft B
PSAR lists both the 2019 and 2023 edition. Similarly, the PSAR states that the RPV will be constructed to the 2023 edition. Neither edition has been endorsed by the NRC. X-energy should consider the questions and update the PSAR accordingly.
79 6
6.4.6 The expression "Level A and B" is used twice in draft PSAR under these sections. Section 6.4.6 C
has the following, for example:
"... the design conditions must bound any condition experienced by the vessel during normal ID Chapter Section Observation Cateaorv operation or expected transients (Level A and B)."
The draft PSAR appears to not define what Levels A and B stand for.
80 6
6.4.6.1 Clarify the order of welding and bolt preloading regarding the top head and vessel lid seal welds.
B The wording in Section 6.4.6.1 is open to interpretation.
X-energy should provide an enlarged views of Figure 6.4.6-6, "RPV Top Head Fuel Inlet," and Figure 6.4.6-11, "RPV Lead Sealing Procedure." Has the potential for cracking the seal welds been considered from loading of the bolts? Clarify the safety significance of the seal welds.
81 6
6.4.6.1 An enlarged view of the seal weld for the control rod drive mechanism housing in Figure 6.4.6-2, B
"RPV Vessel Top Head Design Features," should be useful. Will a butter be applied to the control rod drive mechanism seal welds?
82 6
6.4.6.2 X-energy should clarify in the PSAR if the reactor pressure vessel (RPV) will be subject to an B
irradiation -
ement surveillance program. The NRC staff no.
hat the EOL fluence for the RPV is > ((
)) n/cm2 and the helium inlet temperature is ((
))°C.
83 6
6.4.7.1 The draft PSAR states: "Supports are constructed to ASCE 43. The RCCS is seismic Category 1 C
with associated safe shutdown earthquake."
Suggested edits: Supports are oeRstF1:1oteEI designed to include ASCE 43 seismic design criteria.
The RCCS is seismic Category 1 with associated safe shutdown earthquake.
84 6
6.4.1 0, Descriptions for the RCSS and RPS appear to have some overlap, and both systems are B
6.4.1 3 classified as SR.
It is not clear if the control system for the RCSS is also SR. Based on the instrumentation and control (l&C) White Paper, the RCSS is controlled by the Distributed Control System (DCS) and the Investment Protection System (IPS) that may not be classified as SR (e.g., NSRST). The NRC staff noted that the draft PSAR states: "{l&C controls information for the RCSS will be provided in a future l&C document.} but suggests providing sufficient details in the PSAR that ID Chapter Section Observation Cateaorv make a clear distinction between the RCSS and RPS initiating functions.
85 6
6.4.11 Will the external hazards analysis of the RCCS and its ductwork consider whether RCCS flow B
rate (driven by RCCS air heat-up/buoyancy alone) could be reduced/impeded by outdoor weather conditions or other anomalous wind/weather effects (wind interactions with the auxiliary building itself or other surrounding structure(s), etc.).? That is, could outdoor conditions apply/create differential pressure between the inlet and outlet ducts significant enough to affect RCCS flow? X-energy should consider clarifying this in the PSAR.
86 6
6.4.11, Figure 6.4.1 1-3, "RCCS Top View," shows that the RCCS ductwork passes through the Nuclear B
1.1.4.4.2, Island Auxiliary Building (NIAB) interior space and then continues through the NIAB exterior wall.
6.4.12.2, The RCCS is SR and the NIAB is NSRST.
7.3.1.4 Section 1.1.4.4.2, "Secondary Systems and Components," (bottom of Page 1.1-15) states that the NIAB is effectively only considered NSRST for asset protection and that "none of the systems located within the NIAB are required for safe shutdown of the plant." This seems potentially contradictory. Additionally, Section 6.4.12.2, "Steam Generator (SR Portion) Performance and Operation," states that the SR steam generator isolation valves are located outside the RB, in the NIAB.
The NIAB wall interface with the RCCS should be considered/described in Section 7.3.1.4, "Nuclear Island Auxiliary Building (NIAB)."
87 6
6.4.11.1 The Xe-100 Draft PSAR states that the resulting site dose for four units due to Ar-41 is below the B
NRC's (10 CFR 20.1301 ) TEDE limit to the public of 1.0 mSv/annum and the NRC Dose Limits 0.5 mSv/annum. In addition, the draft PSAR states that Ar-41 was the only radionuclide included in the offsite public dose analysis but provided no other details or references to the anticipated Chapter 9 discussions on effluents. Does the statement about the Ar-41 being the only radionuclide included in the analysis for public dose intend to speak for the radionuclides used as source term in Chapter 9? Or was this referring to only the sources that are seen from RCCS air to the environment?
It may be useful to include references to the discussions in Chapter 9 if there are plans to include ID Chapter Section Observation Cateaorv a more detailed analysis of the normal operational plant effluents in Chapter 9. Within the Chapter 9 discussions, the NRC staff would expect a supplemental discussion on effluents beyond Ar-41 (including those that may come from other sources in the plant), or why Ar-41 is their primary/only contributor.
88 6
6.4.13 The NRC staff suggests that the PSAR include, in addition to some of the figures in B
Section 6.4.13, "Plant Control and Data Acquisition System," and Section 7.3.20, "Plant Control and Data Acquisition System (PCDAS)," an overall l&C system architectural figure (see Figure 18 of the white paper, "Xe-100 Plant Control and Data Acquisition System, Revision 3," which has been discussed with the NRC staff). The figure will be useful for the NRC staff in understanding the overall Xe-100 l&C design architecture and how the fundamental l&C design principles, e.g.,
redundancy, independence, and diversity, are accomplished at a high level. Although preliminary, the figure should show all l&C systems (including safety classification), interfaces, communication with direction of data flow, and isolation devices, and isolation of information being sent off-site (e.g., hardware diode). These are the areas of focus for the NRC staff for the l&C area.
89 6
6.4.13.1.2, Section 6.4.13.1.2, "RPS Performance and Operation," states that the RPS uses set of trips and B
Table logic to initiate protective actions prior to exceeding analytical limits, hence protecting associated 6.4.13.1-2 safety limits. The NRC staff suggests a table that contains information on such protective actions, plant process sensors relied upon, their ranges covering anticipated scenarios, trip setpoints, and analytical limits.
90 6
Table of Chapter 6, Table of Content, lists Section 6.4.13, "Plant Control and Data Acquisition System."
C Contents This system covers l&C subsystems that are not just SR but NSRST and NST (based on the white paper provided to the NRC staff). The overall l&C architecture and interrelationships in the system are important but clarity can be made to the fact that Chapter 6 is intended for SR.
91 6
All How is stress relaxation cracking being accounted for in Alloy 800(H)? RG 1.87, Revision 2 B
applicable specifically states that stress relaxation cracking shall be considered in high-carbon austenitic (face-centered crystal structured) materials at elevated temperatures. X-energy should consider providing a relevant discussion in the PSAR.
ID Chapter Section Observation Cateaorv 92 6
Table The DBHL for seismic events is PGA of ((
]Jg in the table. According to Appendix S to B
6.1.3-1 1 O CFR Part 50, the minimum PGA for the horizontal component of the SSE at the foundation level in the free-field should be O.1 g or higher.
X-energy should explain the value of PGA ((. ]Jg as DBHL with regards to meeting 1 O CFR Part 50 Appendix S requirements.
93 6
Table The 'Summary of Basis for Parameters' column simply lists reference documents or studies for B
6.1.3-1 the DBHLs identified. It should include more specific information regarding how the selected DBHL is justified for the site or design. This information should help the NRC staff understand the clear bases for the DBHLs.
In addition, NEI 21-07 states that: "In Chapter 2 will also serve as the location for other SAR material addressed by ARCAP guidance. This would include summaries of the site-related information and analyses used to develop the DBHLs documented in Section 6.1.1." With the methodologies and analyses section in draft PSAR for Project Long Mott being in Chapter 3, the site-related information and analyses in Chapter 3 may also include similar information regarding the bases for the DBHLs.
94 6
Table 6.2-1 Capability targets for individual SR SSCs are identified in Table 6.2-1, "SR SSC Reliability and B
Capability Targets," (note that the table itself is not numbered). By virtue of being {yellow}, it is understood that validation records are not yet completed. However, Section 6.2.1, "References,"
indicates that there are no references (it states: "None."). Moreover, there is no call out of references anywhere in the section.
- Confirm that there is no intent to provide a validation reference for the capability and reliability targets. If confirmed, provide a rationale for not specifying the source.
- If there is an intent to provide references for these values, please clarify what type of documents will be the source. As an example, for core graphite structures (AOO and DBE cases), the capability is defined in terms of the probability of failure (POF) (for crack initiation) as specified in ASME BPV Section Ill, Division 5 for the applicable service limits. As discussed in several Division 5 subsections (e.g., HHA-3100, HHA-3130, and HHA-3217), the POF is ID Chapter Section Observation Cateaorv calculated using, among other parameters, a reliability curve specified in the Material Data Sheet (HHA-2200). Development of the material data sheet requires qualification testing. Please clarify what type of source document is expected to be the validation source for this reliability value at the CP and operating license (OL) submittal.
95 6
Table 6.2-1 Clarify if the reliability target frequencies in Table 6.2-1 reflect or encompass the capability B
targets at the EOL? It is expected that the probability of graphite crack initiation and growth will rapidly increase above the turnaround fluence.How will graphite cracking be monitored to demonstrate the reliability target frequencies are being met?
96 6
Table 6.2-1 In Table 6.2-1, the preliminary reliability targets (e.g., 1 E-04) for the RPS and other l&C-related B
systems are listed. Can X-energy explain what specific methods and/or codes and standards are used in demonstrating how the SSC meets the reliability target? The NRC staff suggests that the PSAR include some discussion on the subject. The information may be closely related to the set of special treatment provisions applied; however, the linkage between the special treatment provisions and the reliability targets should be explained.
97 6
6.4.1 Section 6.4 of the PSAR provides a description of SR SSCs, including its RSFs and applicable B
design criteria. The NRC staff expects that the PSAR clearly identify the codes and standards that X-energy proposes to use for SR SSCs to provide reasonable assurance that design criteria and RSFs will be met.
X-energy should address the following with regards to proposed codes and standards for SR structures:
(1) Table 6.4.1-2, "Load Combinations for Seismic Category I (Safety-Related) Steel SSCs," in Section 6.4.1.1 presents the load combinations for SR steel structures. However, the notes refer to the code for concrete structures instead of the code for SR steel structures. X-energy should reconcile this inconsistency.
(2) The NRC has endorsed AISC N690-18 in RG 1.243, with exceptions, additions, and clarifications for use in SR steel and steel composite structures. This endorsement includes load combinations acceptable to the NRC staff. X-energy should consider using AISC N690-18 and ID Chapter Section Observation Cateaorv RG 1.243 for the design of SR steel structures.
(3) Section 6.4.1.1, page 27, lists the codes and standards for the design of the RB. X-energy should clarify why there are multiple/duplicative codes listed for the design of SR steel structures.
For example, AISC N690-1994 and its 2004 Supplement 2, AISC N690-1 8 are listed, as well as AISC 360-1 6. The NRC staff notes that AISC now uses N690-18 for SR steel structures. AISC 360 is the parent code of AISC N690-18, which already accepts the provisions in AISC 360 with the additions and exceptions noted in AISC N690-18. X-energy should clarify in the PSAR why AISC 360 is referenced on page 27.
(4) Section 6.4.1.1, page 27, lists the codes and standards for the RB. For concrete structures it listsACI 349-13, which is the standard endorsed in RG 1.142. ltalso listsACI 318-08.
ACI 318-08 is the parent code for ACI 349-13, which accepts its provisions with the exceptions and additions in ACI 349-13. X-energy should clarify in the PSAR, why is it necessary to refer to ACI 318-08.
98 6
Table Table 6.4.11 -1, "Design Values," lists the minimum inlet temperature as ((
)]°C. Is the B
6.4.1 1-1 material of construction used for the RCCS susceptible to a ductile-to-brittle transition temperature? X-energy should consider the question and update the PSAR accordingly.
99 6
Table Table 6.4.12-1, "Xe-1 00 Steam Generator Isolation System Code & Standards B
6.4.12-1 ASME Code Case (it should be ASME/ANSI Standard) N-278.1, "Self-Operated and Power-Operated Safety-Related Valves Functional Specification Standard" is listed.
ASME/ANSI Standard N278.1 -1975 was endorsed in RG 1.148, "Functional Specification for Active Valve Assemblies in Systems Important to Safety in Nuclear Power Plants."
However, RG 1.148 was withdrawn per Federal Register/Volume 75, Number 11/Tuesday, January 19, 2010/Notices, which stated thatASME/ANSI Standard N278.1 -1 975 had been superseded by ASME QME-1. ASME QME-1 defines requirements and provides guidelines for qualifying active mechanical equipment used in nuclear power plants. ASME QME-1-2017 is endorsed in RG 1.100, Revision 4. Can X-energy explain the use of ASME/ANSI Standard ID Chapter Section Observation Cateaorv N278.1-1975 instead of ASME QME-1-2017? X-energy should consider the questions and update the PSAR accordingly.
100 6
Table The Highly Integrated Protection System (HIPS) platform L TR is incorporated by reference in B
6.4.13.1-1 PSAR Section 6.4.13.1.1, "RPS Functions and Design Criteria." Note that other applicants that are also using the HIPS platform are planning to submit a supplement to the HIPS L TR to justify the differences in the HIPS platform design from the design approved in the topical report.
101 6
Table The NRC staff notes that the HIPS L TR incorporated by reference in the PSAR is based on B
6.4.13.2-1 versions of industry consensus standards that are different from the ones outlined in Table 6.4.13.2-1, "Xe-100 Nuclear Instrumentation System Codes & Standards." The NRC staff suggests X-energy to discuss the differences in the PSAR with explanations or justifications.
102 6
Table The draft PSAR states that neutron source pebbles are considered NSRST, despite containing B
6.4.2-1 radionuclides (i.e., source material). Should these pebbles also support RSF 1 and therefore be SR? Additionally, startup pebbles are used to control reactivity during startup. Should they be considered SR on that basis? X-energy should consider the questions and update the PSAR accordingly.
103 7
All Table 7.3.3-1, "SGS Codes and Standards": The title for BPVC-11I-5, "Vessels & Piping," is not C
applicable correct. It should be "Rules for Construction of Nuclear Facility Components - High Temperature Reactors."
Section 7.3.5.1, "Cross Vessel System Functions and Design Criteria," the last paragraph on Page 7.3-56, regarding the discussion of the design conditions for the cross vessel, the sentence
'The Primary Loop Pressure Relief System (PLPRS) valves are required to lift at the design pressure shown in Section 6.4.6.1 Reactor Pressure Vessel Functions and Design Criteria" seems irrelevant. Also, there seems to be an editorial error with cross reference: When Section 6.4.6.1, "Core Graphite Structures and Design Criteria," is referenced, the section title "Reactor Pressure Vessel Functions and Design Criteria," also shows up.
104 7
All sections The NRC staff previously provided feedback on the white paper on Xe-100 Seismic Design C
Methodology. X-energy should consider the feedback and update the PSAR accordingly.
ID Chapter Section Observation Cateaorv 105 7
All sections Some codes and standards in the references are older editions and not endorsed by current B
RGs.
Please update to current codes or explain the use of older codes, vice codes and standards as endorsed by RGs and/or newer editions with improved provisions.
106 7
7.2 Table 7.2-1, "NSRST SSC Special Treatment Requirements," does not contain the needed A
information and is to be filled out later at the level of detail allowable by the current design {yellow bracketed}.
In addition, the NRC staff suggests that the codes and standards in this table be consistent with, or evaluated against, those discussed in individual SSC description sections under Section 7.3, "Descriptions for NSRST SSCs."
107 7
7.3.3 Figure 7.3.3-1, "SG Schematic Diagram," is mentioned but is missing in Section 7.3, C
"Descriptions for NSRST SSCs," of the draft PSAR.
108 7
7.3.5 It states that: "Consistent with PDC RFDC 2, the CV [cross vessel system] is designed to perform C
its safety-significant functions in the event of a safe shutdown earthquake and other natural phenomena hazards."
In NEI 18-04, RFDCs are associated with SR SSCs. Since this system is NSRST (vs SR), did X-energy intend to state PDC 2? Other NSRST SSC sections use PDC 2.
The NRC staff notes that Revision 2 of the PDC topical report has a PDC RFDC 2 expression that may need to be corrected.
X-energy should consider the feedback and update the PSAR.
109 7
7.3.8.3 It states: "Only SR SSCs are required to be qualified for seismic events environmental conditions, B
and electromagnetic compatibility. The helium circulator system is NSRST, so equipment qualification is not required." NSRST SSCs should be qualified for their expected environmental ID Chapter Section Observation Cateaorv conditions during normal operations and licensing basis events. Thus, this statement above appears not justified. The corresponding sections for other systems state the need for appropriate environmental qualification. For example, Section 7.3.4.3, "Hot Gas Duct System,"
states: 'The hot gas duct system is designed, per the identified design codes, for the expected operating environmental conditions." X-energy should consider the comment and update the seismic qualification requirement for the helium circulator as appropriate.
110 7
7.3.14 Table 7.3.14-1 lists the electrical system applicable codes and standards. Please explain how B
IEEE standards for Class 1 E or electrical SR systems are partially applicable to a NSRST electrical system (for example, IEEE Standards 308, 336, 338, 382, 387, 384, 603, 690, and 7 41 ). If they are only used as guidance, please clarify this.
11 1 7
7.3.14 Section 5.3, "Principal Design Criteria," includes the PDCs and includes PDC CDC 17. Please C
provide a discussion on PDC regarding the inspection and testing of electrical systems for clarity.
112 7
7.3.15 This section seems to be incomplete, even on its own terms, as will be seen in some of the other B
observations for this section.
For the CP application, the NRC staff expects to receive a complete section.
113 7
Table The discussion of "Assumption 1" introduces the term "Fire Protection Analysis (FPA)" as distinct C
7.3.15-1 from Fire Hazards Analysis, Fire PRA, and Fire Safe Shutdown Analysis. The term is not defined or described, and it is unclear what it is intended to mean. A clear definition should be provided in the PSAR.
114 7
Table One of the columns is titled "Regulatory Guide 1.189, Revision 4, Regulatory Requirement."
C 7.3.15-1 The Staff Positions documented in RGs are not regulatory requirements, but instead guidance.
115 7
Table The Basis column discusses the applicability of the codes and standards. Some of the B
7.3.5-1 discussions appear to be incorrect or is unclear.
Does X-energy intend to follow them with or without deviations? Since many of them are for SR ID Chapter Section Observation Cateaorv SSCs, does X-energy intend to "grade" its application as appropriate consistent with NSRST SSC classification?
Some codes and standards are approved by the NRC via RGs with exceptions and/or clarifications. For example, RG 1.151 endorses ANSI/ISA-67.02.01-1999 with exceptions and clarifications. Does X-energy intend to follow the RGs?
Instead of the specific codes and standards, the table lists RGs in some cases (e.g., RG 1.153, which endorses IEEE 603-1991 ). Is this a consistency issue?
More specific observations are as follows:
- ANSI/ASME NQA 1 - "Applicable at the organizational/plant level, thus to some degree applies to all systems."
- Explain "NQA 1... to some degree applies to all systems." ASME OM-HTGR - "Section applicable to HTGRs"
- Explain "... applicable to HTGRs"
- IEEE 323 - "Applicable to all SR mechanical and electrical systems." It is only applicable to SR electrical (including l&C) systems.
- IEEE 344 - "Applicable to all SR SSCs." It is only applicable to SR electrical (including l&C) and active mechanical systems.
- ISA 67.02.01 - "Applicable to SR instrument lines"
- ISO 9001 - "Applicable at the organizational/plant level, thus to some degree applies to all systems"
- Explain "ISO 9001... to some degree applies to all systems."
ID Chapter Section Observation Cateaorv
- NUREG/CR-6303 - "Applicable to all SR and NSRST SSCs"
- RG 1.153 - "Applicable to all SR mechanical, electrical, and l&C systems." It is only applicable to SR electrical and l&C systems.
X-energy should consider the questions and update the PSAR accordingly.
116 8
8.1.2 Section 8.1.2, "Systems, Structures, and Components Classifications," of the PSAR states, in C
part, that "NEI 18-04, Table 4-1 relates to quality assurance (QA) requirements to SSC classifications as follows: SR SSCs: QA requirements are consistent with a 10 CFR Appendix B QA program and implemented in a risk-informed and performance-based manner." This is not the exact quote from NEI 18-04, Table 4-1, and as restated by X-energy in the PSAR, seems to suggest that the requirements of Appendix B will be implemented in a risk-informed and performance-based manner, which is not consistent with the regulations. The determination of the applicability of Appendix B can be risk-informed, but its implementation cannot. Please clarify the intent of this statement.
117 8
8.1.2 Related to reliability assurance program (RAP), this section has the following:
A "Additionally, the QA program, along with the NEI 18-04 process, are utilized to meet the guidance described in SRP 17.4 through the development and verification of reliability and capability targets and identification and establishment of special treatments."
RAP, as discussed in NEI 18-04, is one of the basic requirements for all safety-significant SSCs and includes reliability and availability targets for SSCs performing PRA safety functions.
Available guidance includes reliability and integrity management (RIM) program (draft PSAR Section 8.2, "Reliability and Integrity Management Program"), SRP 17.4 (RAP), and SRP 19.3 (RTNSS per SRMs to SECY-94-084 and SECY-95-132).
RAP is needed at the pre-operational stage as one of its objectives is to provide reasonable assurance that a plant is designed, constructed, and operated in a manner that is consistent with the risk insights and key assumptions (e.g., SSC design, reliability, and availability) from the probabilistic, deterministic, and other methods of analysis used to identify and quantify risk.
ID Chapter Section Observation Cateaorv At the operational stage, the maintenance program and QA program can be used for the RAP (as discussed in Standard Review Plan (SRP) 17.4 ).
The NRC staff observes that additional description should be provided for RAP. For example, what elements of the X-energy's NEI 18-04 process address the expectations/acceptance criteria for RAP? How does X-energy intend to meet SRM-SECY-95-132 regarding RAP? X-energy should consider the questions and update the PSAR accordingly.
118 8
8.1.3 Section 8.1.3, "Determination of the QA Program Requirements" states, in part, that "If the SSC C
forms part of a pressure boundary function or performs a pressure-retention function, assess and classify the pressure boundary risk and select an appropriate pressure boundary QA program."
Is the purpose of this assessment to classify the SSC as either SR, NSRST, or NST to determine the applicability of Appendix B to 10 CFR Part 50 and 10 CFR Part 21?
X-energy should consider the questions and update the PSAR accordingly.
119 8
8.2 Section 8.2, "Reliability and Integrity Management Program," states that the Xe-100 RIM program B
is implemented in accordance with ASME Section XI, Division 2, and describes the general RIM process. It does not provide any specific information related to the Xe-100 RIM program.
RG 1.246 endorses the 2019 Edition of ASME Section XI, Division 2, with conditions. Condition 1 of RG 1.246 identifies information that should be submitted with the application and refers to Note 6, which states that: "If information is not complete at the time of application, then the application itself should contain a schedule for when the information will be completed and submitted to the NRC. For example, if the RIM program is being submitted for a construction permit (CP) then such application should describe the information that would be included in an OL application."
The NRC staff encourages that the Xe-100 RIM program follow the conditions specified in RG 1.246. In the CP application, X-energy should describe the RIM information that would be included in an OL application.
120 8
8.3 This section has a brief program discussion.
C ID Chapter Section Observation Cateaorv
'The plant maintenance program controls the performance of maintenance activities in the plant.
It provides assurance that the actual plan configuration is consistent with the capabilities modeled in the licensing base event analyses. The program applies to safety significant SSCs within scope."
The draft PSAR also states that RG 1.160 and NUMARK 93-01, which is endorsed by the RG, are "for LWRs and thus informed the program."
Similar to other operational program discussions under Section 8.4, "IST Program," additional information related to the program description, applicable requirements (i.e., 10 CFR 60.65 with the potential for an exemption request related to a reactor coolant pressure boundary - See ARCAP Roadmap ISG Appendix B), and a program plan may be useful. As a minimum, a statement such as "the maintenance program will be described... in the application for an operating license," as done for the 1ST program, should be added for clarity.
121 8
8.5 Multiple chapters and sections of the draft PSAR for various SR and NSRST SSCs refer to this A
section as the SSCs' environmental qualification program special treatment. However, Section 8.5, "Environmental Qualification Program," does not include a discussion on the environmental qualification program for these safety-significant SSCs. In addition, Chapter 6 and Chapter 7 sections for SR and NSRST SSCs should be reviewed to ensure that their reference to Section 8.5 as special treatment is consistently applied. Some sections do not refer to Section 8.5, although the reference appears to be needed. The NRC staff expects that Section 8.5 discusses the environmental qualification program for all safety-significant (SR and NSRST) SSCs, thus this information needs to be provided in the PSAR.
122 8
8.5 Section 7.3.6.3, "Start-up and Shutdown System Qualifications, Testing, and Programs," states B
that motor-operated valves (MOVs), motors, instrumentation and potentially other equipment are expected to be qualified for harsh environments. In addition to IEC/IEEE Std. 60780-323, what qualification standards will be utilized for these specific equipment (i.e., IEEE has specific qualification standards for motors, actuators, etc.)?
123 8
8.6 Applicability of Commission Policy statements on advanced reactors.
B ID Chapter Section Observation Cateaorv The Commission established enhanced criteria for fire protection and other technical areas through policy statements in the 1990's (SECY-90-016, SECY-93-087, SECY-94-084, and associated SRMs). Chapter 8 of RG 1.189 reflects the Commission's direction. The enhanced criteria simplify fire protection programs for advanced designs by restricting or eliminating the locations containing all trains of a safe shutdown function.
The proposed PSAR sections do not address these criteria, and Table 7.3.15-1, "Compliance of RG 1.189," does not address RG 1.189 Chapter 8 at all.
124 8
8.6 The text describes the Safe Shutdown Analysis as being informed by the Fire PRA.
C The NRC staff notes that significant results of the Safe Shutdown Analysis (i.e., identification of required safe shutdown functions, identification of required SSCs) are inputs to the PRA, rather than results from the PRA.
125 8
8.7.2 Section 8.7.2, "Scope," states that the human factors engineering (HFE) program as proposed by B
NUREG-0711 is divided into 12 elements, including Staffing & Qualifications (S&A). The regulations in 10 CFR 50.54(m) specify minimum control room staffing for LWRs. The NRC Commission Policy for post-TMI actions requires an independent Shift Technical Advisor (STA) position to provide engineering expertise and advice to the shift supervisor in the event of abnormal or accident conditions. The NRC staff expects that operating small modular reactors will seek exemption from the requirements. In the past, the NRC staff used NUREG-1791,
"Guidance for Assessing Exemption Requests from the Nuclear Power Plant Licensed Operator Staffing Requirements Specified in § 50.54(m)," to review staffing exemption.
The HFE Program Plan should include a high-level approach for staffing because the staffing plan is an input to other HFE elements.
The NRC staff suggests that X-energy indicate in Section 8.7, "Human Factors Engineering Program," the plan for a staffing exemption from 50.54(m), whether a STA role is planned to be used or eliminated, and if yes, what approaches/methodologies will be used, and reference NUREG-1791, if X-energy plans to use it.
ID Chapter Section Observation Cateaorv 126 8
8.7.3 Section 8.7.3, "Assumptions and Constraints," states that development of operational experience B
review (OER) will be limited due to plant design differences. NUREG-0711, "Human Factors Engineering Program Review Model," Paragraph (3) in Section 3.4.1 states that: "Related HSI Technology - The applicant's OER should cover operating experience with the proposed HSI technology in the applicant's design."
The following observations are based on NUREG-071 1 and the related supporting guidance:
Even without similar types of plants in operation, OER should be extended to surrogating systems that share similar concepts of operation (e.g., operating multi-module units) and human-system interaction characteristics (e.g., automation, advanced displays and alarms) to those of Xe-100 design.
In addition, since the Xe-100 design will be highly automated, OER should include consideration of any automation failures that required human actions and human errors in using automation.
127 8
8.7.6 PSAR Section 8.7.6, "HSls, Procedures, and Training," indicates that a graded approach is B
applied to the analysis of human-system-interface (HSI) design. Since the design philosophy introduces concepts of operation different from light-water reactor plants, some HSI design principles are expected to be different. For example, one fundamental human factors principle is that operators are always in control of HSI automation (references: NUREG-0700, Chapter 8, and IEEE-1786 guidance for computerized procedures). This may not be true with new designs.
So, a high-level analysis of HSI design is tied to Functional Requirement Analysis (FRA) and Functional Allocation (FA) analysis. This should be described in the HSI analysis. The NRC's guidance for HSI design, NUREG/CR-0700, Chapter 9 describes some principles and criteria for HSI automation design. Does X-energy intend to consider NUREG/CR-0700 Chapter 9 in Functional Allocation (FA) and the early stage of the HSI analysis? If yes, this should be mentioned in the PSAR section.
128 9
All sections No discussions or placeholders provided for the discussion of PDC that affect the design of A
waste systems. Within the ARCAP Chapter 9 guidance, there are a few places where the acceptance criteria requests the applicant to describe the PDC that affect the design of the liquid, ID Chapter Section Observation Cateaorv gaseous, and solid waste systems. If there are no PDC that influence the design of these systems, please state so. Given some of the preliminary nature of the PDC review, this should be addressed in some manner for the official submittal. Discussions related to PDCs 60, 61, 63 and 64 tied into Chapter 9 or references made to Chapter 7 discussions for the applicable PDC is expected.
129 9
9.1 In Section 9.1, "Gaseous Waste Management," the applicant provides reference to Section 7.3.9, B
"Helium Service System," for the description of the helium service system (HSS). This reference is useful since it describes the system used to monitor the helium for the facility. However, this reference only provides for the monitoring and treatment of helium within the facility. Does the facility anticipate the need for treatment of other gaseous releases in the facility? Are any other anticipated gaseous sources outside of the noble gases swept from the primary coolant?
Anticipation of any activated argon? How does the facility's ventilation system handle the monitoring of air outside of the HSS? This should be mentioned in the PSAR.
130 9
9.1, 9.2, For ARCAP Chapter 9, the NRC staff did not observe any discussions or placeholder discussions A
9.3 related to addressing RG 1.143 design criteria besides the applicability to mobile or temporary radwaste treatments systems. The NRC staff also did not notice anything contained within Section 7.3.9, "Helium Service System," Section 7.3.12, "NI Liquid Radwaste System," or Section 7.3.13, "Radioactive Waste Treatment Facility", for the system design criteria either. Is this discussion contained elsewhere? A reference to where the application addresses the RG 1.143 system classifications for the radwaste treatment systems would be useful.
131 9
9.1.3.1 Section 9.1.3.1, "Discharge Requirements and Release Points," provides information on the B
monitored discharge points. Are there any figures or high-level drawings that can help with understanding the plant environments that these radiation effluent monitors will monitor. The NRC staff would like to better understand the flow paths of effluent and potential effluent from the sources to the monitoring locations summarized in Section 9.1.3.1. This information would be used to determine the potential radionuclides released, then ensure that appropriate monitoring is available to support the detection of the anticipated effluents.
132 9
9.2 In Section 9.2, "Liquid Waste Management," the tables or references to information that describe C
the capacity of tanks are useful for addressing the facility's ability to store and process liquid ID Chapter Section Observation Cateaorv waste. This information is useful for the PSAR when comparing the anticipated liquid waste generation rates from various sources and ensuring adequate capacity to allow for storage and treatment of liquid waste. During the review of Table 7.3.13-1, "RWT Equipment Design Information," some tank capacities are provided. References to this table along with any other would be useful in assessing liquid waste storage capacities.
133 9
9.3.2 In Section 9.3.2, System Description, for the PSAR review, the NRC staff is looking to B
understand the amount of solid waste storage space that will be available in the radwaste building. Based on the amount of solid waste storage available, how long will it take for the building to be filled? How long does the design envision storing solid waste? For the OL application, the expectation is that more of the durations of storage will be addressed, but as part of the initial design it is important to ensure adequate storage space will be available for storage of waste.
134 10 All sections For discussing the applicant's radiation protection organization and training of workers, ARCAP C
Chapter 10 guidance states that in accordance with 10 CFR 50.34(a)(6) the PSAR for a CP needs only to provide the following information related to the control of occupational dose: a preliminary plan for the applicant's organization, training of personnel, and conduct of operations.
In addition, a CP application should include a description of its plans to develop, at the OL or stage, comprehensive worker protection programs, organizational structure, training, and monitoring to ensure the requirements in 1 O CFR Part 19 and 1 O CFR Part 20 are met.
It is recognized that some of this information is contained in Chapter 11. The NRC staff suggests providing pointers within Chapter 10 to state where the information to satisfy the ARCAP Chapter 1 O information requests is located.
135 10 10.2 In Section 10.2, "Radiation Sources," X-energy provides a format that appears to indicate it will B
provide information on its sources in the PSAR. The NRC staff also notes that in addition to the source term information for these sources, other parameters such as assumptions for model volumes, material thicknesses, and assumed material densities are information that would be requested to facilitate any sort of NRC staff confirmation of subsequent dose assessments in Chapter 10.
ID Chapter Section Observation Cateaorv 136 10 10.3.1.3 Definitions such as those requested below should be provided in the PSAR. Details on the A
radiation zoning details may or may not be provided in the PSAR.
In Section 10.3.1.3, "Radiation Zoning and Access Control," X-energy indicates that more details for the radiation zones will be provided in the PSAR, as well as providing access controls in the PSAR. The NRC staff notes that it will be important to state what the access controls are and provide what the definitions will be for the planned zoning criteria (ex. what Radiation Areas, High Radiation Areas, Very High Radiation Areas, etc., definitions will be). Also, the NRC staff expects to see what controls to restrict access are provided for specific zones such as discussing how compliance with 10 CFR 20.1601 and 10 CFR 20.1602 will be achieved.
137 10 10.3.2 ARCAP Chapter 1 O states: "... a description of the design-specific PDC necessary to control C
occupational exposures should be provided in the CP application."
In Section 10.3.2, "Shielding," the applicant provides input on how the shielding design will support PDCs 4, 19, and 61. While the NRC staff observes that the design bases writeup reads as if it addresses the PDC, it may be useful to clearly state the connection to the PDC and how the specific design of shielding meets each of the PDC because of the information stated by the design objectives.
138 10 10.4.1 This detail not normally expected in the PSAR. If this information is provided, the following C
conveys the expectations for content:
In Section 10.4.1, "Occupational Radiation Exposure," the applicant provides information stating that there is no separate determination of doses due to airborne activity planned and this determination is based on experience demonstrating that dose from airborne activity is not a significant contributor to total doses. This is a difficult concept to work through since it would be dependent on the location of work and the types of work being performed in the area. The NRC staff would expect that to support this statement the applicant would demonstrate based on calculations that the expected airborne activity is far below the limits set in 10 CFR Part 20. This could look like a radiation area map for expected airborne concentrations. While this information may not be provided in the application, it could be subject to audit review to verify the statements made by the applicant.
ID Chapter Section Observation Cateaorv 139 10 10.4.2 This detail not normally expected in the PSAR. If this information is provided, the following C
conveys the expectations for content. In Section 10.4.2, "Public Dose Assessment," the applicant indicates that direct radiation from its facility is indistinguishable from background based on the plant radiation zoning requirements. To support this the NRC staff would be looking to either audit or review information available in the application to verify information such as how large sources of radiation within the facility impact other area dose rates and to verify the plant radiation zoning requirements are established correctly. This would subsequently verify the statements that direct radiation sources to the public are indistinguishable from background sources.
140 10 10.5 Preliminary information on the applicant's plans to establish an operational radiation protection C
plan are useful but not required in the PSAR.
In Section 10.5, "Operational Radiation Protection Plan," the NRC staff notes that this is likely an OL stage input. References made to NEI 07-03A and NEI 07-08A are useful if they support the type of radiation protection program envisioned by this applicant. These references are also discussed in ARCAP guidance and in the roadmap guidance.
141 8
8.7 Section 8.7, "Human Factors Engineering Program," references NUREG-0711, Revision 3, but C
does not reference NUREG-0700. Does X-energy intend on following guidance for NUREG-0700?
Note: this observation, while addressing Chapter 8, was originated during the review of Chapter 11 and located here for that reason.
142 11 11.2 Organizational structure defines education and experience requirements, and Reference 11.2-3 B
cites RG 1.8, Revision 4. Does X-energy intend to follow the endorsed guidance of RG 1.8?
143 11 11.2 Reference 11.2-2 is RG 1.33, Revision 3, "Quality Assurance Program Requirements C
(Operation)". The submittal does not reference the RG in the PSAR. While the NRC staff recognizes that X-energy committed to RG 1.33 in its L TR on quality assurance programs, RG 1.33 is not mentioned in PSAR Chapter 11.
ID Chapter Section Observation Cateaorv 144 11 11.2, Regulation 1 O CFR 50.34(a)(6) states that a CP application must include a preliminary plan for B
11.2.4.1 the applicant's organization, training of personnel, and conduct of operations. The NRC staff notes that the draft PSAR contains a construction organizational chart that shows maintenance groups supporting the construction of the facility, along with training, but these groups are absent in the operational organizational chart. Also, it is not clear how is maintenance support of the operational unit being managed or supervised? In addition, it is not clear how is the training group being managed or supervised?
145 11 11.2.2.5 DANU-ISG-2022-05 Acceptance criterion i: "The applicant has described the role and function of A
the architect engineer and the nuclear steam supply system vendors during design and construction, as well as the organizational controls over the project-related activities of the architect-engineer and nuclear reactor vendors, including preservation of documentation."
The submittal does not include a description of the preservation of documentation.
146 11 11.2.4.1 Section 11.2.4.1, "Operating Organization Structure," does not provide information on support A
groups to the operating organization. For example, what is the organizational group responsible for maintaining procedures and managing the corrective action program?
147 11 11.3 In Reference 11.3-2, the Agencywide Documents Access and Management System (ADAMS)
C accession number is inaccurate. The number "ML2310A063" is missing a number 7; it should read "ML23107A063."
148 11 11.4 This detail not normally expected in the PSAR. Does X-energy plan to use a plant reference or a C
commission-approved simulator for operator licensing examinations?
149 11 11.4.1 The submittal does not reference NEI 06-13A or any other guidance related to cold-plant B
operator licensing. Does X-energy intend to follow the NRC endorsed guidance of NEl-06-13A?
150 11 11.7 This detail not normally expected in the PSAR. If this information is provided, the following C
conveys the expectations for content:
ID Chapter Section Observation Cateaorv It is agreed that the submittal does not require a physical security or cybersecurity plans for the CP application. However, the Table of Exemptions for LIC-116 lists exemptions regarding regulations for these security programs. Early engagement with NRC staff on the details of these exemptions would be beneficial for the licensing process.
151 11 11.7 The list of references is rather short and excludes major foundational security-related guidance C
(e.g., NUREG-0800 13.6 and 13.6.1 and 13.6.3, RG 5.54, RG 5.69, RG 5.76, and NEI 03-12, etc.). Will a more exhaustive list be consulted to prepare the OL? X-energy should consider the questions and update the PSAR accordingly.
152 11 11.7.1 Regarding the following statement "... A more detailed description of Physical Security Plan to be C
submitted with OL."
(1) Will the detailed Physical Security Plan itself be submitted with the OL as required by 10 CFR 50.34(c) "Physical Security Plan"?
(2) What will be the scope/composition of the security organization given the related proposed exemption discussed in the "Table of Exemptions"?
X-energy should consider the questions and update the PSAR accordingly.
153 11 11.7.3 Regarding the following statement "... A more detailed description of Safeguards Contingency C
Plan to be submitted with OL."
Will the detailed Safeguards Contingency Plan itself be submitted with the OL, as required by 10 CFR 50.34(d) "Safeguards contingency plan."
154 11 11.7.5 X-energy has elected to subject individuals specified in 10 CFR 26.4 to the provisions in B
10 CFR 26 Subpart K for the Xe-100 plant.
'The Xe-100 FFD Program for construction adheres to the guidance in NEI 06-06, "Fitness for Duty Guidance for New Nuclear Power Plant Construction Sites" [Reference 11.7-3].
ID Chapter Section Observation Cateaorv The NRC endorsed NEI 06-06, with exception, in RG 5.84, "Fitness for Duty Programs at New Reactor Construction Sites" [Reference 11.7-4]."
Does X-energy intend to commit to RG 5.84, Revision 0, July 2015, which endorses, with exception, NEI 06-06, Revision 6, April 2013?
NEI 06-06 covers individuals described in 10 CFR 26.4(f), but other individuals would be subject to an FFD program during construction (like those individuals in 10 CFR 26.4(e) and (g)). These individuals would be subject to 10 CFR Part 26, subparts A-H, N and 0. X-energy should consider reviewing NUREG-0800, Chapter 13.7.2, Table 1, "FFD Program Applicability and Milestones."
X-energy should consider the questions/observations and update the PSAR accordingly.
155 11 11.7 Regarding the following exemptions:
B Regulation 1 O CFR 73.55(e)(5) - Bullet resistance for the main control room (MCR) (Operators do not perform SR functions).
Regulation 1 O CFR 73.55(k)(5) - Requirement for armed on-site security guards (DBT can't result in exceeding dose limits).
Is this an exhaustive list of exemptions? Can the applicant provide additional insight into the proposed exemptions as identified, including justification and timing of its submittal. It would appear that the 10 CFR 73.55(e)(5) MCR exemption could be submitted at the CP stage given MCR construction considerations.
Based on the Table of Exemptions, is the applicant considering exemptions related to Part 53 or Limited Scope for Security rulemakings (not approved)?
X-energy should consider the questions and update the PSAR accordingly.
156 Part V All sections The way this document is structured is confusing in that each page has a header of 'Table of C
ID Chapter Section Observation Cateaorv Exemptions" even in Table V.2-1, "List of Non-Applicabilities for the Xe-1 00," which lists regulation subsections that X-energy identifies as not applying to its design (and which therefore will not need exemptions). X-energy should provide clarification.
157 Part V V.2 Table V.2-1 lists 10 CFR 50.34(f)(2)(x) and (f)(2)(xi), related to relief and safety valves, as not B
applicable due to the HSS not causing core cooling challenges. The draft PSAR includes the operation of HPB relief valves as part of the LBE discussions. In addition, Section 3.3.9, "Small SG Tube Leak (1-1 0mm)," discusses a safety relief valve lift within the reactor vessels and subsequent radiological release to the environment. Has X-energy evaluated the applicability of the regulations against the safety relief valves and HPB relief valves, which may not be part of the HSS?
X-energy should consider the questions and update the PSAR accordingly.
158 Part V V.3 X-energy proposes an exemption to replace the use of "important to safety" with "safety B
significant SSCs".
In RG 1.189, "important to safety" is defined as:
Nuclear power plant SSCs "important to safety" are those required to provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public. In Appendix R to 1 0 CFR Part 50, "important to safety" and "safety related" apply to all safety functions.
For the Fire Protection Program, X-energy should address the differences between "safety significant" vs. "important to safety" and "safety related" in the context of the RG 1.189 definition.