ML25071A178

Nuscale, Response to SDAA Audit Question Number: A-5.4.3-4

ML25071A178
Person / Time
Site:	05200050
Issue date:	03/12/2025
From:	NuScale
To:	Office of Nuclear Reactor Regulation
Shared Package
ML25071A176	List: ML25071A177 ML25071A178
References
LO-180268
Download: ML25071A178 (1)

Text

Response to SDAA Audit Question Question Number: A-5.4.3-4 Receipt Date: 04/08/2024 Question:

FSAR Section 5.4.3 states the decay heat removal system (DHRS) removes post-reactor trip residual and core decay heat and transitions the NPM to safe shutdown conditions without reliance on electrical power or operator action. A safe shutdown condition is one in which reactor subcriticality, decay heat removal, and radioactive material containment are properly maintained for the long term. SECY-94-084 provides that relaxed acceptance criteria can be used by passive plant designs to demonstrate that GDC 34 is met for residual heat removal system capability.

The staff performed an audit (ML23067A300) of the DHRS thermal-hydraulic calculation that supports FSAR statements regarding the ability of the US460 DHRS to achieve and maintain a safe shutdown condition. The staff observed that the engineering calculation indicates that

(( 2(a),(c) In addition, during the audit NuScale stated that the ((

}}

2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

Information Requested NuScale is requested to clarify the purpose and use of the (( }} 2(a),(c) The staff requests NuScale to clarify the decay heat removal requirements of the DHRS and (( }} 2(a),(c) should include consideration for single and two-train operation, as well as off-nominal conditions. Corresponding changes should be made to relevant documents (e.g., FSAR, non-LOCA topical report). Additionally, the staff requests that FSAR Section 5.4.3.3 be revised to remove the word active in the first sentence to align with the regulatory requirements established in GDC 34.

Response

Engineering calculation EC-101197, Revision 4, DHRS Thermal Hydraulic Calculation for NPM-20, addresses a corrective action to more accurately model the ((

}}2(a),(c) and is in the electronic reading room (eRR). Figure 5.4-9 through Figure 5.4-12, Figure 5.4-14, and Figure 5.4-15 are updated in the attached markups to the Standard Design Approval Application (SDAA) Final Safety Analysis Report (FSAR) to incorporate the EC-101197, Revision 4, results.

NRC Feedback from January 17, 2025: The updated response to A-5.4.3-4 states that the Chapter 5 results currently presented in FSAR Revision 1 are conservative compared to those using the new NRELAP5 basemodel and therefore FSAR Chapter 5 will not be updated. The staff understands NuScales position that it considers the analytical results in the FSAR conservative, since they illustrate a reduced DHRS capability compared to the current calculation. However, the staff notes that the new basemodel contains numerous changes that impact the response of the DHRS, resulting in a changed NuScale Nonproprietary NuScale Nonproprietary

cooling capacity response. Based on the staffs review to-date, it observes that, while most of the DHRS modeling changes are realistic, these changes are non-conservative in compared to the previous modeling used because these changes can cause the analytical results to gain margin. For this reason, the modeling changes are considered non-conservative. Accordingly, the staff cannot reach the same conclusion that that leaving the FSAR description without an update to the new calculation is conservative. Due to these model changes, the DHRS performance results presented in FSAR Chapter 5 are outdated and are no longer representative of the base modeling on which the FSAR is based. Because the new basemodel contains non-conservative changes and is a materially different representation of the design that credits aspects of the design different than is currently presented in FSAR Revision 1, Section 5.4.3, NuScale is requested to update all the Chapter 5 FSAR Figures for DHRS performance using the new NRELAP5 basemodel. Additionally, the FSAR text description should be updated with the credited heat removal surface area used, which is consistent with the basemodel and DHRS design. Response to January 17, 2025 NRC Feedback: The attached markups to the SDAA FSAR Section 5.4.3 include a statement about decay heat removal system (DHRS) headers and piping below the pool water level contributing to DHRS heat transfer. Chapter 5 of the SDAA FSAR does not include specific details about modeling; specific heat transfer modeling approaches in NRELAP5 are proprietary information. The previous calculation revision did credit heat transfer from piping below the pool water level contributing to DHRS heat transfer. Figure 5.4-9 through Figure 5.4-12, Figure 5.4-14, and Figure 5.4-15 are updated in the attached markups along with corresponding results in SDAA FSAR Section 5.4.3 to incorporate the EC-101197, Revision 3, results. There is currently an open condition report (CR) on EC-101197, Revision 3, that is being evaluated in NuScales corrective action process (CAP). The CR is specifically investigating ((

}}2(a),(c) and a condition evaluation and extent of condition have been assigned.

Action Item from December 11, 2024: NuScale needs to update DHRS performance calc or provide justification for why not. Response to Action Item: NuScale updated the DHRS calculation to evaluate the impact of the new NRELAP5 basemodel on the results. The results in Chapter 5 are conservative compared to those using the new NuScale Nonproprietary NuScale Nonproprietary

NRELAP5 basemodel; therefore, the DHRS performance information in the FSAR Chapter 5 will not be updated. Text in FSAR Section 5.4.3.3.4, Thermal-Hydraulic Performance, including FSAR Table 5.4-5, Decay Heat Removal System Design Data, is still correct and requires no changes. Table 1 below shows the FSAR figures and the corresponding updated figures in EC-101197, Revision 3, DHRS Thermal Hydraulic Calculation for NPM-20, for comparison. The figures in the FSAR show conservative heat removal for the DHRS compared to results using the new basemodel in EC-101197, Revision 3, and no revision is necessary. Engineering calculation EC-101197, Revision 3, is posted in the eRR. Table 1: Comparison of Final Safety Analysis Report Figures and Decay Heat Removal System Thermal Hydraulic Calculation Results with New NRELAP5 Basemodel FSAR Figure Number FSAR Figure Title EC-101197, Revision 3, Figure Number EC-101197, Revision 3, Figure Title 5.4-9 Primary Coolant Temperature with Decay Heat Removal System Single Train 4-12 RCS Temps from Nominal Transient Case, with One Train of DHRS 5.4-10 Primary Coolant Temperature with Decay Heat Removal System Two Trains 4-11 RCS Temps from Nominal Transient Case, with Two Trains of DHRS 5.4-11 Primary Coolant Temperature with Decay Heat Removal System Two Trains High Inventory 4-17 RCS Temps from High Inventory Case, with Two Trains of DHRS 5.4-12 Primary Coolant Temperature with Decay Heat Removal System Two Trains Low Inventory 4-14 RCS Temps from Low Inventory Case, with Two Trains of DHRS 5.4-14 Primary Coolant Temperature with Decay Heat Removal System One Train High Inventory 4-18 RCS Temps from High Inventory Case, with One Train of DHRS 5.4-15 Primary Coolant Temperature with Decay Heat Removal System One Train Low Inventory 4-15 RCS Temps from Low Inventory Case, with One Train of DHRS In addition, there was discussion of nodalization of the pool in the EC-101197, Revision 2, calculation during the meeting with the NRC on December 4, 2024. (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

Executive Summary The NRC staffs concerns about DHRS modeling are due to a misunderstanding of the use of the NRELAP5 code to analyze the performance of DHRS to remove decay heat in the NuScale Power Module (NPM-20); the use of the different models was explained in a clarification call on September 11, 2024. (( }}2(a),(c) The simplified DHRS model is not used to calculate results shown in the Final Safety Analysis Report (FSAR). Chapter 5 of the FSAR demonstrates single-train and two-train DHRS operation and its ability to remove long-term decay heat. Chapter 15 of the FSAR demonstrates that the non-loss-of-coolant accident (non-LOCA) event acceptance is met using the non-LOCA NRELAP5 evaluation model (EM). The DHRS operation considers single active and passive failures, as shown in FSAR Table 5.4-8, Failure Modes and Effects Analysis - Decay Heat Removal System. The NRC staffs concerns (( }}2(a),(c) are due to a staff misunderstanding about the difference between normal operating transients and anticipated operational occurrences (AOOs). Response to October 25, 2024 Feedback: In response to feedback received on October 25, 2024, SDAA Section 5.2.2.2.1 addresses normal operating transients, not AOOs. The staff cited the following statement from FSAR Section 5.2.2.2.1: Sizing of the RCS and the PZR steam space avoids an RSV lift during normal operational transients that produce the highest RPV pressure at full power conditions. The term normal operational transients in the cited statement does not include AOOs. Design sizing calculations demonstrate that the system or component response to normal operating transients do not result in reactor trip or actuation of other safety systems when normal support systems are functional. Normal operating transients are the American Society of Mechanical Engineers (ASME) Service Level A transients defined in FSAR Section 3.9.1.1, Design Transients. From Section 3.9.1.1: Service Level A includes conditions associated with events that are planned to occur due to routine operation of the plant. Examples include startup, power maneuvers, hot shutdown, and shutdown. NuScale Nonproprietary NuScale Nonproprietary

The Service Level A transient definitions for the NPM-20 is in ER-101144, Revision 3, Pressure and Thermal Transient Definitions for Analysis of NSSS Components, which is in the SDAA Audit Chapter 3 (except 3.7 and 3.8) eRR. Anticipated operational occurrences are equivalent to ASME Service Level B transients as defined in FSAR Section 3.9.1.1 and are not considered to be normal operational transients. Evaluation of AOOs is in FSAR Chapter 15. Chapter 15 analyses contain conservative assumptions and assume that the transients progress in the worst possible manner with the goal of ensuring Chapter 15 specific acceptance criteria are met. For example, during heatup events in FSAR Section 15.2, Decrease in Heat Removal by the Secondary System, where RSV lifts are shown to occur, the turbine bypass system is assumed to be unavailable as a conservatism. Operation of the turbine bypass system as expected during these events would play a significant role in preventing RSV lifts. These analyses also do not credit pressurizer spray for reducing pressure and use conservatively biased valve signal and actuation times that contribute to higher pressurizer pressures. Chapter 15 analyses assume the RSVs are open if it is conservative for a particular transient progression. Service Level A transients do not assume this and follow guidance in the Standard Review Plan for Section 3.9.1 and the ASME Code. Using NRELAP5 input for Chapter 15 analyses will yield different results than using the NRELAP5 input for DHRS sizing from EC-101197, Revision 2. The design of DHRS and the RSV are correctly described in the SDAA Chapter 5. The assumptions in FSAR Chapter 15 are inherently different and lead to significantly different results. The statement in FSAR Chapter 5 does not imply conclusions about the results of FSAR Chapter 15 analyses. The staffs statement that normal operating transients includes AOOs is incorrect. The NRC staffs feedback also indicates a concern that (( }}2(a),(c) Technical Specification Limiting Condition for Operation 3.5.4 address operability requirements for the ESB and includes a verification of the ESB after operations that could affect the form or quantity of boron in the ESB dissolvers. ((

}}2(a),(c) Therefore, the technical specification for ESB already addresses the concern identified in the NRC staffs feedback.

Feedback received on August 15, 2024: The aspects associated with items 2 through 4 are considered resolved. The staff provides the following feedback related to item 1. NuScale Nonproprietary NuScale Nonproprietary

The audit response states the (( }}2(a),(c) (( }}2(a),(c) The criteria and performance demonstration should include consideration for single and two-train and off-nominal conditions. Additionally, any impacted documents (e.g., FSAR, non-LOCA LTR, etc.) should be updated to provide this clarification. ((

}}2(a),(c)

Revised Response posted to Chapter 5 eRR on September 24, 2024: (( }}2(a),(c) Long-term stable or decreasing core average temperature and pressure following DHRS actuation in response to a design-basis event demonstrate that decay heat is NuScale Nonproprietary NuScale Nonproprietary

being removed, either solely by DHRS or by DHRS in conjunction with other passive safety features. The discussion in TR-0516-49416, Section 4.3.4.2, is consistent with the safe, stabilized condition identified in SDAA Section 15.0.4. Section 5.4 of the SDAA contains DHRS performance results and addresses single-train nominal, two-train nominal, two-train off-nominal with multiple inventory and heat transfer options, and single-train off-nominal DHRS operation to demonstrate that the system is appropriately sized to remove decay heat and reach a safe stable condition in the event of a single failure, consistent with SECY 94-084. Figure 5.4-11, Figure 5.4-12, Figure 5.4-14, and Figure 5.4-15 in SDAA show the results of the DHRS sizing analyses showing that a safe stable condition is met following a single failure; the figures are based on calculations and NRELAP5 runs in EC-101197, Revision 2, which is in the Chapter 5 eRR for staff review. The NRELAP5 runs in EC-101197, Revision 2, (( }}2(a),(c) Figure 5.4-11, Figure 5.4-12, Figure 5.4-14, and Figure 5.4-15 in the SDAA represent DHRS transient responses ((

}}2(a),(c)

Section 5.4.3.3.4 of the SDAA discusses the different models used; the simplified model represents the steady-state cases, which determine the impact of the performance factors discussed above. Table 5.4-8 referenced in Section 5.4.3.3 considers both single active and passive failures and demonstrates that the DHRS is single-failure proof. Attached markups remove text from Section 5.4.3.3. NuScale Nonproprietary NuScale Nonproprietary

Original Question: The decay heat removal system (DHRS) design basis description in FSAR Section 5.4.3.1 states: The DHRS provides cooling for design basis events when normal secondary-side cooling is unavailable or otherwise not utilized. The DHRS removes post-reactor trip residual and core decay heat from operating conditions and transitions the NPM to safe shutdown conditions without reliance on electrical power or operator action. For passive designs, including the NuScale NPM-20, the safety-related function is to reach a safe shutdown condition and comply with 10 CFR 50, Appendix A, General Design Criteria 34 [or PDC 34] following a non-LOCA event, which is defined as the capability to reduce the RCS temperature to less than 420F within 36 hours. EC-101197, DHRS Thermal Hydraulic Calculation for NPM-20, assesses the capability of the DHRS to bring the NMP to safe shutdown conditions under nominal and bounding, off-nominal conditions. This engineering calculation used (( }}2(a),(c) The staff has several areas of concern related to the justifications provided to date: 1). Neither the non-LOCA LTR methodology nor EC-101197 includes the final IFR designs to be used in the NPM-20plant. It is also not clear what IFR design was used within EC-101197, and whether that input parameter is being tracked as an open design issue. Without including justification for the NRELAP5 DHRS model in the Non-LOCA LTR which addresses different figures of merit from that assessed in EC-101197 and the final IFR design, it is unclear whether the NRELAP5 models that are used to assess DHRS performance and figures of merit for Chapter 5 safe shutdown have been demonstrated to be able to conservatively predict the DHRS heat removal, given the fact that the models differ from the design (for example, but not limited to, the final IFR design) and validation basis provided for the non-LOCA topical report and corresponding non-LOCA figures of merit. 2). The NRELAP5 code was benchmarked against NIST-2 DHRS test data. (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

((

}}2(a),(c) (( }}2(a),(c) ((

}}2(a),(c) To resolve the above deficiencies related to the demonstration of DHRS heat removal capability, NuScale is requested to either: 1) Provide further justification of the NRELAP5 DHRS model addressing the items listed above, or 2) Develop additional models or modifications, to conservatively predict the minimum decay heat removal and demonstrate the current DHRS design is adequate to meet GDC 34 with its application to safe shutdown requirements, or 3) Propose a COL item for a COL applicant to demonstrate GDC 34 is met for DHRS heat removal Original Response: Based on the identification of the question as relevant to FSAR Section 5.4.3, the focus of the question on General Design Criterion (GDC) 34 / Principal Design Criterion (PDC) 34 NuScale Nonproprietary NuScale Nonproprietary

compliance, and the discussion of the non-LOCA topical report and EC-101197, Revision 2, NuScale understands the concern indicated by this question to be related to demonstrating DHRS heat transfer capacity. The concern indicated by this question is not related to structural integrity of components during DHRS operation. The response addresses the question in context of the concern about DHRS heat transfer capacity. NuScale agrees that a fundamental safety-related functional requirement of the NuScale Power Module (NPM) DHRS, is to conform to PDC 34. The FSAR Section 3.1.4.5 outlines the requirement: A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded. Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to ensure that the system safety function can be accomplished, assuming a single failure. There is no requirement in either PDC 34 or in 10 CFR 50, Appendix A, GDC 34 to reduce the reactor coolant system (RCS) temperature to less than 420 degrees F within 36 hours; cooling to less than 420 degrees F within 36 hours only appears in SECY 94-084. Section 5.4.3.3.4 of the FSAR states the general performance recommendations in SECY 94-084 are to have sufficient capacity with passive decay heat removal systems to reduce the RCS temperature to 420 degrees F within 36 hours and that reaching a safe, stable condition is possible in the event of a single failure. Section 5.4.3.3.4 goes on to outline how the design considers the SECY 94-084 recommendations. Calculation EC-101197, Revision 2, DHRS Thermal Hydraulic Calculation for NPM-20, is the basis for the analysis of DHRS heat removal capacity in Section 5.4.3.3.4. Calculation EC-101197, Revision 2, is in the eRR with this response. The areas of concern the staff has in the first numbered list are addressed in the following subsections. Item 1) In the response to request for supplemental information (RSI) 14 (submitted by NuScale letter LO-145424 dated July 14, 2023 and available in ADAMS at ML23195A091), NuScale provided a detailed explanation of the impact of the SG tube inlet flow restrictor (IFR) on various NuScale Nonproprietary NuScale Nonproprietary

analyses. The response included evaluation of the effect of the IFR during DHRS operation, including:

(( }}2(a),(c)

(( }}2(a),(c)

(( }}2(a),(c)

(( }}2(a),(c)

(( }}2(a),(c) While the NPM-20 IFR design is undergoing a change based on in-process DWO analyses, the attached markup shows inclusion of the wide range of Kinlet values considered for the US460 NuScale Nonproprietary NuScale Nonproprietary

design in Section 5.4.1.3 of the FSAR. Because the Kinlet value is the means of modeling the IFR in NRELAP5, no other final IFR design details are pertinent to this work. This item indicates concern about the basis for the NRELAP5 models used to evaluate the DHRS performance and figures of merit for Chapter 5 safe shutdown, and whether they have been demonstrated to be able to conservatively predict DHRS heat removal, considering effects of the IFR design and different figures of merit for Chapter 5 compared to non-LOCA event figures of merit. The IFR design is addressed above. The statement that the model differs from the design in other areas is not clear. As identified in the audit question, with respect to the figures of merit for the analyses presented in Chapter 5, the concern is demonstrating adequately conservative prediction of DHRS heat removal with NRELAP5 as applied in the models documented in EC-101197, Revision 2. The application of NRELAP5 in EC-101197, Revision 2, relies on validation performed originally to support the non-LOCA EM. The application of NRELAP5 in EC-101197, Revision 2, is appropriate to predict DHRS heat removal because:

(( }}2(a),(c)

(( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

The audit question states The applicant (( }}2(a),(c) However, the applicability of the NRELAP5 model for analysis of DHRS cooling is based on separate effects tests and integral effects tests, as summarized in response to item (2) below.

(( }}2(a),(c)

(( }}2(a),(c)

The plant transient analysis in EC-101197, Revision 2, is (( }}2(a),(c) The analysis addressed the increase in core power and therefore, higher decay heat, of the US460 design to evaluate the capacity of the DHRS to provide adequate cooling under nominal and off-nominal conditions. Therefore, the application of NRELAP5 in EC-101197, Revision 2, is appropriate to predict DHRS heat removal to demonstrate adequate cooling capacity, considering the SECY 94-084 recommendations. Item 2) It is important to note that the capability of the NRELAP5 code to model behavior in the SG and DHRS loop during DHRS operation was previously validated during the NRC review and approval of the US600 design and associated methodology topical reports. The non-LOCA topical report (TR-0516-49416-P, Revision 4) Section 5.1.4 includes subsections for each high-ranked phenomenon that identifies how that phenomenon was addressed from an EM validation perspective. As identified in Section 5.1.4, the validation basis is extensive and includes assessments to test data from: NuScale Nonproprietary NuScale Nonproprietary

(( }}2(a),(c) The NPM design changes and changes in operating ranges between the US460 design and the US600 design were evaluated, and it was concluded that ((

}}2(a),(c) Therefore, the existing EM validation basis remains applicable. As described in Section 5.3.7.2.1 of the non-loss-of-coolant accident (non-LOCA) topical report, TR-0516-49416-P, Revision 4, the objective of the NIST-2 DHRS testing was to enhance the previously-approved DHRS validation database.

((

}}2(a),(c) are discussed in EE-107403, Revision 2, and in ER-126485, Revision 0 (these documents have been previously provided in the eRR for the non-LOCA topical report). (( 
}}2(a),(c)

(( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

(( }}2(a),(c) (( }}2(a),(c) (( }}2(a),(c) (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary

((

}}2(a),(c)

((

}}2(a),(c)

Item 3) ((

}}2(a),(c) Any future NRELAP5 analyses will use version 1.7.

(( }}2(a),(c) It is not necessary to reanalyze DHRS results in Chapter 5 with NRELAP5 version 1.7 based on the ((

}}2(a),(c)

NuScale Nonproprietary NuScale Nonproprietary

Item 4) (( }}2(a),(c) The response to Item 2) above describes that ((

}}2(a),(c)

Conclusion The audit question ended with the NRC staffs identification of three possible resolution paths. In this response, NuScale has provided further justification of the NRELAP5 DHRS model addressing the specific areas of concern, consistent with the first listed option. Markups of the affected changes, as described in the response, are provided below: NuScale Nonproprietary NuScale Nonproprietary

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-6 Draft Revision 2 Audit Question A-5.1.3.4-1, Audit Question A-5.4-1 The SG supports and SG tube supports provide support for vibration and seismic loads. As shown in Figure 5.4-5, the SG tube supports and tube support assemblies attach to upper SG supports welded to the PZRintegral steam baffle plate and inner surface of the RPV, and also interface with lower SG supports welded to the inner surface of the RPV. The use of eight sets of tube supports assemblies limit the unsupported tube lengths, which ensures the SG tubes do not experience unacceptable flow-induced vibration (FIV). Figure 5.4-1 shows two of the eight sets of tube supports. As shown in Figure 5.4-5, the lower SG supports permit thermal growth and provide lateral support of the tube supports. Inlet Flow Restrictors The SG inlet flow restrictors are installed in each SG tube at the FW plena locations. Each SG inlet flow restrictor is individually installed and seats against the secondary face of the FW plenum tubesheet and extends into a portion of the hydraulically expanded SG tube within the FW plenum tubesheet. A SG inlet flow restrictor consists of a mandrel, an expanding collet, a flanged sleeve, a locking plate and a hex nut. After the flow restrictor is inserted into the SG tube, the metallic collet on each SG inlet flow restrictor is expanded to seal with the inner diameter of the SG tube. The bearing contact resistance between the expanded collet and tube prevents bypass flow around the flow restrictor as well as the frictional interaction for securing the flow restrictor within the FW plenum. Secondary side water flows from the FW plenum through a center-flow orifice in the mandrel. The flanged sleeve allows secondary side water from the feed plenum to enter into the space between the sleeve and SG tube to FW plenum tubesheet weld. This secondary side water provides a thermal barrier to the tube-to-tubesheet weld, helping mitigate rapid temperature changes at the weld. The devices permit in service tube inspections, cleaning, tube plugging, repairs and maintenance activities via installation and removal as needed. Audit Question A-5.4.3-4 The SG inlet flow restrictors provide a nominal loss coefficient between 800 and 2000. This range corresponds to the SG tube flow area and the Reynolds number of the operating steam generators. Thermal Relief Valves A single thermal relief valve is on each FW line upstream of the tee that supplies the feed plenums (Figure 5.4-7). The thermal relief valves provide overpressure protection during shutdown conditions for the secondary side of the SGs, FW and steam piping inside containment, and the DHRS when the secondary system is water solid for SG flushing operations and the containment isolation system is actuated. The trapped fluid is subject to heating by core decay heat. The thermal relief valves are spring operated relief valves that vent directly to containment. The thermal relief valves are classified as Seismic Category I and Quality Group B

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-19 Draft Revision 2 system components. Table 5.4-4 identifies the component materials used in the DHRS design. Table 5.4-5 provides a summary of DHRS design data. Before power operations, the FW pumps fill the FW lines, SGs, and DHRS. Maintaining filled and pressurized DHRS passive condensers and piping occurs by connection to the FWS on the DHRS outlet line to the FW piping inside containment. During normal power operations, the DHRS is in a standby configuration with each train of DHRS isolated from the associated MS lines by the closed DHRSAVs. These four valves, two in parallel on each train, remain closed. Automatic actuation of the DHRS occurs using the MPS and has the capability for manual initiation from the main control room. The DHRS actuation signal opens the DHRSAVs for both trains of DHRS and closes the secondary system isolation valves (FWIV, feedwater regulating valve (FWRV), MSIV, and secondary MSIV). The MPS automatically actuates the DHRS. Manual controls for initiation of DHRS are also provided in the main control room. Details of the DHRS actuation design including redundancy, reliability, diversity, signals, interlocks, analytical limits, and functional logic are provided in Chapter 7. Audit Question A-5.4.3-4 Upon actuation, the MSIVs and FWIVs close, and the DHRSAVs open. The DHRSAVs open upon interruption of control power because of control system actuation or loss of power. The DHRSAVs use the same hydraulic system used for the CIVs. Section 6.2.4, Containment Isolation System, contains a discussion of hydraulic system operation. Actuation permits the water column in the DHRS piping to drain into the FWS piping and plenum, and steam to flow from the SG into the DHRS piping and the DHRS passive condenser. Steam condenses in the passive condenser by transfer of heat to the reactor pool. The DHRS headers and piping below the pool water level also contribute to the DHRS heat transfer. This process results in a flow of condensate from the passive condenser to the associated FW line and into the associated SG. Figure 5.4-7 depicts the system layout and interface with the MS and FW piping and SGs. The DHRS function depends on the closure of the associated safety-related MSIVs and FWIVs. In the event an MSIV fails to close, the backup MSIV provides isolation for the DHRS loop. The FWRV provides isolation in the case where the FWIV fails to close. These closures isolate the SGs and associated DHRS loops from the MSS and FWS, ensuring adequate water inventory in the passive closed loop configuration. Section 6.2.4, Containment Isolation System, and Chapter 7, Instrumentation and Controls, describe the MSIV and FWIV functions, including their actuation. The SGs are described in Section 5.4.1. Chapter 10, Steam and Power Conversion System, describes the MS and FW piping. Natural circulation resulting from the density differences between the steam and condensate portions of the DHRS and associated SG drive DHRS flow. The DHRS passive condensers are at a higher elevation relative to the SGs to promote natural circulation flow to the SGs. The RCS temperature and pressure

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-23 Draft Revision 2 During DHRS operation, condensate line pressure indication yields a pressure close to the saturation pressure at the RCS temperature. Consistent with steam pressure, higher pressure may indicate a SG tube failure and a lower pressure may indicate a DHRS break or leak. Decay Heat Removal System Controls The DHRS control system is limited to an on-or-off signal to the actuation valves with no ability for modulation. The DHRS actuates from the MPS, as discussed in Chapter 7. 5.4.3.3 Performance Evaluation Audit Question A-5.4.3-4 The DHRS provides the passive, safety-related, single active failure-proof, and redundant capability to cool the reactor core and coolant to safe shutdown conditions. Both liquid and vapor water are in the DHRS on system actuation. The total water mass remains constant during system operation because the DHRS is a closed system. Two independent trains of passive cooling loops ensure reliability of the DHRS. Table 5.4-8 provides the failure modes and effects analysis for the DHRS. The DHRS piping and passive condensers are around the exterior of the CNV and separated to reduce the potential for a single condition to affect both trains. Submergence of the passive condensers in the reactor pool and the module bay walls located between operating NPMs, as shown in Figure 1.2-5, and Figure 1.2-6, provides protection from adverse interactions with other facility equipment. The RXB crane is used to move NPMs to and from the refueling area as discussed in Section 9.1, Fuel Storage and Handling, provides protection from adverse interaction with an NPM being moved to and from the refueling location. Section 9.1.4, Fuel Handling Equipment, describes refueling and maintenance operations conducted in the refueling area. The DHRS is not functional or available during refueling operations. 5.4.3.3.1 Water Hammer Loading conditions due to water hammer in the DHRS or surrounding systems are included in the analysis of the DHRS.The operating conditions for the main FWS, MSS, and DHRS lines are conducive to water hammer events caused by high pressure discharge. fast valve closure. pump trip transients. The FW piping operates at a much higher pressure than atmospheric pressure and operates in such a way that prevents column rejoining from occurring.

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-27 Draft Revision 2 of loop leakage from high inventory cases creates the most conservative limiting heat transfer case. Audit Question A-5.4.3-4, Audit Question A-5.4.3.3.4-1 SECY 94-084 provides general performance recommendations for passive decay heat removal systems to have sufficient capacity to reduce the RCS temperature to 420 degrees F (safe shutdown condition) within 36 hours and that reaching a safe stable condition is possible in the event of a single failure. SECY 94-084 was considered in the development of the DHRS design capacity with an alternate acceptance criteria of achieving a passively cooled, safe shutdown condition. For the US460 design, the technical specifications define the conditions for passive cooling, which include single-train operation of DHRS to account for a failure that removes the functionality of an entire train. Cases are evaluated for single-train and two-train operation at nominal initial conditions, both of which show that the DHRS is capable of bringing the NPM to a passively-cooled safe shutdown condition. The consideration of potential DHRS failures as initiating events is addressed in Chapter 15. Decay Heat Removal System Performance Results The system performance analysis indicates the DHRS removes appreciable amounts of heat over a wide range of initial conditions. Audit Question A-5.4.3-4 Figure 5.4-9 shows RCS cooldown for 36 hours from full power conditions with one DHRS train in operation assuming nominal system conditions. Initially, the decay heat exceeds the combined heat removal of the DHRS. The decay heat power drops off quickly as the transient progresses, and the DHRS begins to remove more heat than is added. This imbalance cools the RCS. This case also demonstrates that a single train of DHRS can provide sufficient cooling of RCS using nominal system conditions to below 400 degrees F RCS average temperature within 36 hours. Audit Question A-5.4.3-4 Figure 5.4-10 shows RCS cooldown for 36 hours from full power conditions with two DHRS trains in operation assuming nominal system conditions. For this nominal two DHRS train case, DHRS provides sufficient cooling toRCS average temperature stabilizes below 300 degrees F RCS average temperature within 36 hours. Audit Question A-5.4.3-2, Audit Question A-5.4.3-4 Figure 5.4-11 shows an off-nominal DHRS actuation with high DHRS inventory and low DHRS heat transfer. The heat removal bias is lower because of the high fouling, high tube plugging, and a high volume of non-condensible gas. This case assumes 102 percent reactor power. For this off-nominal two DHRS train case, DHRS provides sufficient cooling toRCS average temperature stabilizes below 42400 degrees F RCS average temperature within 36 hours.

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-28 Draft Revision 2 Audit Question A-5.4.3-2, Audit Question A-5.4.3-4 Figure 5.4-12 shows an off-nominal DHRS actuation with low DHRS inventory and low DHRS heat transfer. The heat removal bias is lower because of the high fouling, high tube plugging, and a high volume of non-condensible gas. This case also uses the presence of loop leakage to further bias results. This case assumes 102 percent reactor power. This event also considers the presence of decreasing inventory due to loop leakage. For this off-nominal two DHRS train case, DHRS provides sufficient cooling to RCS average temperature stabilizes below 350420 degrees F RCS average temperature within 36 hours. Audit Question A-5.4.3-2, Audit Question A-5.4.3-4 Figure 5.4-14 shows an off-nominal DHRS actuation with high DHRS inventory and low DHRS heat transfer with one train of DHRS active. The heat removal bias is lower because of the fouling, tube plugging, and volume of non-condensible gas. This case assumes 102 percent reactor power. For this off-nominal one DHRS train case, DHRS provides sufficient cooling to below 450 degrees F RCS average temperature within 36 hours and shows the RCS passive cooling is maintained. Audit Question A-5.4.3-2, Audit Question A-5.4.3-4 Figure 5.4-15 shows an off-nominal DHRS actuation with low DHRS inventory and low DHRS heat transfer with one train of DHRS active. The heat removal bias is lower because of the fouling, tube plugging, and volume of non-condensible gas. This case uses the presence of loop leakage to further bias results and assumes 102 percent reactor power. This case also considers the presence of decreasing inventory due to loop leakage. For this off-nominal one DHRS train case, DHRS provides sufficient cooling to below 400 degrees F RCS average temperature within 36 hours, with a trend showing that the temperature will continue to decrease. Audit Question A-5.4.3-2, Audit Question A-5.4.3-4 The final results show that the DHRS is capable of removing appreciable amounts of heat over a relatively wide range of inventories. The analyses further show the ability to accommodate fouling, SG tube plugging, and the presence of noncondensible gas, thus precluding the need for high-point vent capability. The transient plots provided in Figure 5.4-11, Figure 5.4-12, Figure 5.4-14, and Figure 5.4-15Figure 5.4-12 include these factors and show that even with the degraded heat transfer, the system meets its requirements. Under each of the off-nominal transients the DHRS provides continuous passive cooling of the RCS. These results confirm that the DHRS is capable of bringing the NPM to a passively-cooled safe shutdown condition, within a reasonable period of time, and with no offsite power or operator action required.

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-61 Draft Revision 2 Audit Question A-5.4.3-4 Figure 5.4-9: Primary Coolant Temperature with Decay Heat Removal System Single Train 350 400 450 500 550 600 650 0 20000 40000 60000 80000 100000 120000 140000 Temperature (°F) Time (sec) Riser Cold Leg Average

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-62 Draft Revision 2 Audit Question A-5.4.3-4 Figure 5.4-10: Primary Coolant Temperature with Decay Heat Removal System Two Trains 250 300 350 400 450 500 550 600 650 0 20000 40000 60000 80000 100000 120000 140000 Temperature (°F) Time (sec) Riser Cold Leg Average

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-63 Draft Revision 2 Audit Question A-5.4.3-4 Figure 5.4-11: Primary Coolant Temperature with Decay Heat Removal System Two Trains High Inventory 350 400 450 500 550 600 650 0 20000 40000 60000 80000 100000 120000 140000 Temperature (°F) Time (sec) Riser Cold Leg Average

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-64 Draft Revision 2 Audit Question A-5.4.3-4 Figure 5.4-12: Primary Coolant Temperature with Decay Heat Removal System Two Trains Low Inventory 300 350 400 450 500 550 600 650 0 20000 40000 60000 80000 100000 120000 140000 Temperature (°F) Time (sec) Riser Cold Leg Average

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-66 Draft Revision 2 Audit Question A-5.4.3-2, Audit Question A-5.4.3-4 Figure 5.4-14: Primary Coolant Temperature with Decay Heat Removal System One Train High Inventory 420 440 460 480 500 520 540 560 580 600 620 0 20000 40000 60000 80000 100000 120000 140000 Temperature (°F) Time (sec) Riser Cold Leg Average

NuScale Final Safety Analysis Report Reactor Coolant System Component and Subsystem Design NuScale US460 SDAA 5.4-67 Draft Revision 2 Audit Question A-5.4.3-2, Audit Question A-5.4.3-4 Figure 5.4-15: Primary Coolant Temperature with Decay Heat Removal System One Train Low Inventory 350 400 450 500 550 600 650 0 20000 40000 60000 80000 100000 120000 140000 Temperature (°F) Time (sec) Riser Cold Leg Average

Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 25 exceeds a predetermined setpoint. The ESFAS coincident logic is designed so that no single failure can prevent a safeguards actuation when required, and no failure in a single measurement channel can generate an unnecessary safeguards actuation. Transients requiring decay heat removal are addressed by the DHRS, which provides cooling through one or both of the SGs. For a steam generator tube failure (SGTF), main steam line break, and feedwater line break, the affected SG is isolated and the DHRS provides cooling through the intact SG (depending on the break location DHRS may be operational in both SGs). Manual operation of the nonsafety-related CVCS can also be used to offset decreases in RCS inventory. If the CVCS is inadequate to address the inventory decrease, containment isolation occurs and the DHRS is actuated. If RCS inventory loss to containment persists, ECCS is actuated. Module-specific systems and functions that operate to mitigate the effects of postulated non-LOCA events (and credited in the safety analysis) include the ECCS, DHRS, CVCS and demineralized water system isolation, MPS, RTS, containment isolation and PZR heater isolation. The only safety system shared between modules is the UHS. The non-LOCA safety analyses consider ranges of UHS conditions and heat transfer such that non-LOCA analysis of a single module bounds possible UHS interactions between modules. 3.3 Decay Heat Removal System Audit Question A-5.4.3-4 The DHRS is a closed-loop, two-phase natural circulation cooling system. Two trains of decay heat removal equipment are provided, one attached to each SG loop. Each train is capable of removing 100 percent of thedesigned to remove decay heat load and cooling the RCS. Each train has a passive condenser immersed in the reactor pool. Upon receipt of an actuation signal, the main steam isolation valves (MSIVs) and the feedwater isolation valves close, and the decay heat removal actuation valves open, allowing heat removal via the SGs. The decay heat removal actuation valves would open upon the loss of power, thus enabling reliable long term cooling. For successful operation, liquid water enters the SG through the feedwater line and is boiled by heat from the RCS. The vapor exits the SG through the steam line and is directed to the DHRS condenser where it condenses back to liquid to return to the SG. Thus, the loop transfers heat from the RCS to the DHRS fluid using the SG and then from the DHRS to the reactor pool water. 3.4 Emergency Core Cooling System The ECCS consists of two or three independent reactor vent valves (RVVs), depending on NPM design, and two independent reactor recirculation valves (RRVs). The ECCS is initiated by simultaneously actuating the RVVs on the top of the RPV in the pressurizer region and the RRVs on the side of the RPV in the downcomer region. Opening the ECCS valves allows a natural circulation path to be established - water is vaporized in the core, leaves as steam through the RVVs, condenses and collects in the containment, and returns to the downcomer region inside the RPV through the RRVs. During normal operation, each ECCS valve is held closed by the hydraulic pressure across the valve main disc. Included in the RRV design is an inadvertent actuation block (IAB) consisting}}