ML24348A008
| ML24348A008 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 12/13/2024 |
| From: | NuScale |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24348A006 | List:
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| References | |
| LO-176318 | |
| Download: ML24348A008 (1) | |
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Response to SDAA Audit Question Question Number: A-15.6.5-1 Receipt Date: 03/27/2023 Question:
The ECCS Low Riser Water Level sensor is a thermal dispersion switch assembly which measures the localized two-phase level in the riser during a LOCA. The LOCA TR, the subsequent LOCA analyses in SDAA section 15.6.5 and ECCS evaluation use collapsed riser water level in the NRELAP5 model. During a LOCA, because of the two-phase level swell and uneven coolant distribution in riser, the measured two-phase level by the thermal dispersion switch assembly could possibly be higher than the collapsed water level and misrepresent the actual coolant inventory. As the result of this potential difference, the actual ECCS actuation could be later than its current design and the impact could be nonconservative. Sufficient technical information and rationales have not been provided to evaluate the impact on the analyses.
Response
The original response to this audit question was provided on April 24, 2023. A supplemental response (including the information beginning with the Response Supplement section) was provided on February 1, 2024. A meeting to discuss this topic was held on August 27, 2024. A copy of the presentation material from the meeting is provided in the electronic reading room (eRR). A second supplemental response was provided on September 5, 2024. The NRC provided feedback on the second supplemental response in October 2024 and a meeting to discuss the NRC feedback was held on October 16, 2024. As a result of the meeting, NuScale is supplementing the response again to respond to the NRC feedback (section titled Supplemental Response to Address NRC Feedback from October 2024) and modify the previously-provided markups. An additional supplement to the response is added to address NRC feedback received in a clarification call held on November 26, 2024 (section titled Supplemental Response to Address NRC Clarification Call on November 26, 2024).
NuScale Nonproprietary NuScale Nonproprietary
The SDAA technical report "NuScale Instrument Setpoint Methodology," TR-122844, identifies various reactor protection system and engineered safety features actuation system setpoints.
The analytical limit for the low RPV riser level ECCS setpoint is specified in Tables 5-20 and 6-1 as having upper and lower limit values of 552 in. and 540 in., respectively. Similarly, the analytical limit for the low-low RPV riser level ECCS setpoint is specified in Tables 5-21 and 6-1 as having upper and lower limit values of 472 in. and 460 in., respectively. The specification of these two analytical limits as ranges is consistent with both SDAA FSAR Tables 7.1-4 and 15.0-
- 7. As identified in the footnote for SDAA FSAR Table 15.0-7, a range is used to accommodate instrumentation uncertainty. In contrast, SDAA FSAR Table 6.3-1 provides singular values for these two ECCS actuation setpoints. While the values are within the range identified in the other tables in the FSAR and technical report, Table 6.3-1 is nevertheless incorrect. To ensure consistency and eliminate confusion, FSAR Table 6.3-1 is updated to specify the same range as identified in the other tables.
The audit request states that the LOCA analyses in SDAA Section 15.6.5 and the ECCS evaluation use collapsed riser water level in the NRELAP5 model. This statement is incorrect. In the LOCA transient analyses using NRELAP5, in both SDAA Chapters 6 and 15, the ECCS actuations associated with RPV riser level are based on void fraction in a cell in order to account for mixture level. Void fractions greater than 90% are used to actuate ECCS, consistent with the description in Section 5.2 of the SDAA topical report Loss-of-Coolant Accident Evaluation Model, TR-0516-49422-P, Revision 3. ((2(a),(c) As will be discussed below, results for MCHFR, minimum collapsed liquid level, or containment response figures of merit are either not sensitive to the ECCS actuation setpoint or the low-low riser level scenarios are non-limiting. NuScale Nonproprietary NuScale Nonproprietary
For evaluation of MCHFR during LOCA in Chapter 15, the limiting MCHFR occurs during Phase 0, which is defined as the time prior to ECCS actuation (Figure 4-1 of the LOCA topical report TR-0516-49422). Therefore, there is no impact of the assumed ECCS actuation setpoint on MCHFR. The potential impacts on collapsed liquid level and containment response are discussed separately below. The LOCA spectrum analyses performed for Chapter 15 primarily use the upper limit value (i.e., 552 in.) for the low riser level ECCS setpoint. However, sensitivity studies are included that also use the lower limit value (i.e., 540 in.) for the low riser level ECCS setpoint. The performance of these sensitivity studies is consistent with the discussion in Section 5.2 of the LOCA topical report TR-0516-49422 that states that the effect of the valve opening time on the event progression is considered. Sensitivity analysis results from the LOCA spectrum analyses are shown in Figures 1 and 2 of this response. In each figure, a solid line represents the transient parameter for a specific break size with an ECCS actuation setpoint of 552 in. and a dashed line of that same color represents the transient parameter for the same break size but with an ECCS actuation setpoint of 540 in. Figure 2 also includes superimposed lines identifying the approximate ranges of the two different riser level ECCS setpoints. The sensitivity studies show that the liquid space breaks and larger vapor breaks (i.e., greater than 35% high point vent break size) actuate ECCS on low riser level regardless of whether 552 in. or 540 in. is used because they result in a relatively rapid depressurization. The small breaks (i.e., less than 35% high point vent break size) have a slower event progression and bypass the low riser level setpoint when 540 in. is used due to the temperature interlock; these smaller breaks actuate ECCS on the low-low riser level signal instead. The resulting differences in ECCS actuation timing can be seen in Figures 1 and 2. The LOCA spectrum sensitivity calculations show that peak containment pressure (Figure 1) is generally higher when the upper limit value of 552 in. is used, although the difference is small for the cases with the largest peak pressure. Cases where use of 540 in. results in delay of ECCS actuation until the low-low riser level have much lower peak containment pressures as shown in Figure 1. The sensitivity calculations also show that the impact on peak containment temperature is negligible, except that cases that actuate ECCS on low-low riser level have much lower peak containment temperatures. Figure 2 shows that using either upper or lower end of the low riser level actuation range, or shifting actuation from the low riser level to the low-low riser level, has a negligible impact on the minimum collapsed liquid level that occurs after ECCS valve opening and RPV blowdown in Phase 1b. The LOCA analysis results presented in FSAR Table 15.6-13 reflect the limiting sensitivity cases, which include consideration of the variation in ECCS actuations. For containment response analyses performed for Chapter 6, the LOCA topical report TR-0516-49422 Table 5-11 identifies that the ECCS actuation ranges are evaluated. Table 5-11 is NuScale Nonproprietary NuScale Nonproprietary
intended to ensure that the analyses using the LOCA methodology consider the range of analytical limits. Consistent with this methodology and the analytical limits specified in SDAA FSAR Tables 7.1-4 and 15.0-7, the LOCA analyses for containment response consider both a high bias (upper end of the analytical range) and a low bias (lower end of the analytical range) for ECCS actuation. The SDAA calculations demonstrate that earlier actuation of ECCS is more limiting since it minimizes the potential for DHRS heat removal prior to blowdown through the RVVs when ECCS actuates. The SDAA calculations also demonstrate that larger break sizes are the most limiting. For these larger breaks, ECCS actuation occurs on low riser level regardless of whether the low riser level is biased to the upper or lower limit. Figure 3 shows the results of discharge line sensitivity calculations which include consideration of low riser level setpoint (i.e., 552 in. vs. 540 in.). Cases DL-1 and DL-4 in Figure 3 are identical except that DL-1 uses a setpoint of 552 in. while DL-4 uses a setpoint of 540 in. The difference in peak pressure between the two cases is negligible (less than 5 psi) compared to the available margin to the acceptance criteria (greater than 200 psi based on FSAR Tables 6.2-1 and 6.2-2). FSAR Section 6.2.1.1.3 explicitly identifies that the limiting containment response results occur for ECCS actuation on low RPV riser level, biased to the high end of the range (i.e., 552 in.). The limiting case corresponds to DL-2 in Figure 3. For extended passive ECCS cooling analyses and evaluation of long-term ECCS performance in SDAA FSAR Section 15.0.5, specific ECCS actuation levels are not studied. Instead, a range of initiating events are evaluated that transition to extended ECCS cooling over a wide set of times. A spectrum of LOCA break sizes is evaluated. Non-LOCA events may also transition to ECCS cooling, on timer actuation 8 hours after reactor trip, to ensure boron transport from containment into the RPV. Given the wide range of conditions considered as entry to the long-term ECCS cooling, and the insensitivity of collapsed liquid level shown in the LOCA break spectrum analysis, no additional sensitivity analyses on ECCS actuation setpoints were performed for the extend passive cooling phase. With respect to the incorrect audit request statement that the LOCA analyses in SDAA Section 15.6.5 and the ECCS evaluation use collapsed riser water level in the NRELAP5 model, it is possible that the confusion results from some of the information in the LOCA topical report TR-0516-49422 regarding previous testing. ((
}}2(a),(c)
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(( }}2(a),(c) Overall, the simulations of Run 2 and Run 5, summarized in Section 7.5.10.7 of the LOCA topical report TR-0516-49422, show that NRELAP5 can reasonably predict the response of the system considering a high point vent line break with different methods and timing of ECCS actuation. The testing concludes that NRELAP5 can reasonably predict the system behavior for a range of ECCS actuations. By actuating ECCS on void fraction in NRELAP5 for the SDAA calculations, and varying the setpoint bias (high or low) depending on the analysis figure of merit, the FSAR Chapters 6 and 15 analysis results address the NRC concern about ECCS actuation. As described previously, FSAR Table 6.3-1 is updated to indicate the setpoint range for consistency with the other tables in the FSAR and instrument setpoint technical report TR-122844. Response Supplement NRC staff feedback on this audit response was provided on January 10, 2024 and the feedback was subsequently discussed on a clarification call between NuScale and the NRC on January 22, 2024. As a follow-up action from the clarification call, NuScale provides the following information to supplement the original response:
A figure showing the relationship between the NRELAP5 node boundaries, modeled ECCS actuation signals, and ECCS actuation setpoint limits is provided in the electronic reading room (eRR) as NPM20 NRELAP5 Nodes with ECCS Actuation Information. The figure is the same as shown during the January 22, 2024 clarification call.
FSAR Section 7.2 is revised (see attached markup) to enhance the descriptions of the RPV riser level measurement consistent with the discussion during the January 22, 2024 clarification call.
Testing performed for the level sensors is provided in the eRR as ER-140431, Rev. A, Proof of Concept Testing of Level Measurement Technology for the NuScale Power Module. (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary
(( }}2(a),(c) In addition, it is important to note that this testing represents one phase of development to demonstrate that a prototypical sensor can meet basic performance requirements. Future phases are identified for further development and eventual qualification of the instrumentation for use in the NPM.
The basic concept of an individual sensor (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary
(( }}2(a),(c) Based on the original response to this audit question, the verbal information provided by NuScale during the clarification call, the supplemental information above added to this audit question, and the additional documents provided in the eRR, NuScale concludes the following:
Operating plants currently use level detection systems with similar technology as the system proposed by NuScale.
Prototypical sensors can meet basic NuScale performance requirements.
Full design and qualification of sensors to support use in an NPM is a future activity that is not required at this stage.
High void fraction at a discrete location is a reasonable surrogate for the method of level detection used by the sensors.
NRELAP5 is validated to accurately predict axial void fraction during LOCA conditions.
Chapter 15 figures of merit are not sensitive to variations in level used for ECCS actuation. Therefore, the existing Chapter 15 safety analyses are adequate to demonstrate reasonable assurance that relevant acceptance criteria are met based on the NPM level detection system and setpoints. NuScale Nonproprietary NuScale Nonproprietary
(( }}2(a),(c) Figure 1: Chapter 15 ECCS Actuation Setpoint Sensitivity Study - Containment Pressure (solid lines are 552 in., dashed lines are 540 in.) NuScale Nonproprietary NuScale Nonproprietary
(( }}2(a),(c) Figure 2: Chapter 15 ECCS Actuation Setpoint Sensitivity Study - Level above Top of Active Fuel (solid lines are 552 in., dashed lines are 540 in.) NuScale Nonproprietary NuScale Nonproprietary
(( }}2(a),(c) Figure 3: Chapter 6 ECCS Actuation Setpoint Sensitivity Study - Containment Pressure (DL-1 is 552 in., DL-4 is 540 in., other DL-# cases are for sensitivities to other parameters) NuScale Nonproprietary NuScale Nonproprietary
Supplemental Response to Address NRC Feedback from October 2024 In October 2024, the NRC provided feedback on NuScales supplemental response dated September 5, 2024 in the eRR. The NRC feedback is provided below in indented, italicized text, with NuScales response following in unindented, regular text. LOCA Topical Report: Unlike the version of the LTR approved for the DCA, Revision 3 of the LTR does not include a listing of the analytical limits. Because the limits are not defined in the method, the staff cannot make a finding for the topical report regarding the adequacy of the modeling approach used to simulate detection of low riser water level and resultant ECCS actuation. The staff intends to develop an L/C to clarify that the approach for modeling riser level setpoint actuation based on detection of a mixture level will need to be provided and justified by an applicant referencing the LTR. TR-0516-49422-P, Revision 3, Loss-of-Coolant Accident Evaluation Model, is revised as shown in the attached markup to provide additional details of the modeling approach. The additional details identify how ECCS actuation is modeled to occur for a given nodalization when applied to a specific design that identifies the analytical limits. It is not necessary to specify the analytical limits themselves in TR-0516-49422-P because they are values associated with a specific design (i.e., an input). With the changes provided in the markup to TR-0516-49422-P, it is not necessary for the NRC staff to develop a limitation and condition for TR-0516-49422-P related to this issue. FSAR: The staff will ensure the FSAR provides sufficient information detailing how the entire range of the analytical limit is modeled, and justification has been provided for the selected mixture level that is simulated for ECCS actuation for each of the LOCA cases. Feedback to NuScale: Based on the approach described above the FSAR is missing: 1) a description of how mixture levels are modeled at the low analytical limit; 2) justification for the selected simulated mixture level actuation setpoint; and 3) the identification of the specific node or cell in the LOCA model that is used to capture the void fraction that corresponds to the mixture level. To address item 1, the staff requests NuScale to describe in the FSAR how the low range is modeled in NRELAP5 (i.e., 540 inches for low level, and 460 inches for low-low level). NuScale Nonproprietary NuScale Nonproprietary
The FSAR Section 15.6.5 discussion in the previously-provided markup includes how the mixture level is modeled corresponding to an approximate value in the range of the analytical limit for low riser level. The FSAR Section 15.6.5 discussion in the previously-provided markup includes how the mixture level is modeled corresponding to an approximate value slightly above the range of the analytical limit for low-low riser level. The markup is revised to add the discussion of the ECCS actuation modeling sensitivity studies for the low range of the low riser level signal (i.e., simulating actuation below 540 inches) that support the conclusion that results are not sensitive to the modeling, or later ECCS actuation is non-limiting. The details of the sensitivity studies are provided in the original audit question response and can be docketed with the response.(( }}2(a),(c) To address item 2, the FSAR should describe the sensitivity analysis that was performed, and the conclusions drawn from it. The staff also requests that the sensitivity results be provided on the docket (NuScale confirm this information is already included in the audit response). FSAR Sections 15.6.5 and 15.6.6 are revised as shown in the attached markup to add a summary of the sensitivity studies performed and the major conclusions. The details of the sensitivity studies are provided in the original audit question response and can be docketed with the response. A summary of key parameters associated with the sensitivity studies is provided in Table 1 below for reference. NuScale Nonproprietary NuScale Nonproprietary
Table 1: Summary of Key Parameters of Emergency Core Cooling System Actuation Modeling Sensitivity (( }}2(a),(c) To address item 3, this clarification can be provided either in the FSAR or in a docketed response (NuScale confirm this information is already included in the audit response). The specific node numbers and elevations are NuScale proprietary information and so are not included in the FSAR. The values are provided in the original audit question response, as well as in Table 1 above, and can be docketed with the response. Proposed changes to the FSAR Sections 7.2.16 & 15.6.5, and Tables 7.1-4 & Table 15.0-7 do provide additional specificity for the RPV riser level sensors performance requirements and acceptance criteria that can be verified during EQ testing, with the following additions in red requested:
Part 2, FSAR Section 7.2.16.4, Sensor Selection (requested additions in red text) The RPV riser level measurement provides input to MPS to actuate ECCS within the specified level range during service conditions typical of when ECCS actuation is NuScale Nonproprietary NuScale Nonproprietary
expected. The RPV riser level instrumentation is qualified to actuate ECCS within specified analytical limits for the spectrum of postulated design-basis loss-of-coolant accidents process conditions (liquid and vapor) and the associated range of depressurizations. Additionally, qualification confirms the acceptability of the correlation between RPV riser level transient and the modeling approach used in the transient and accident analyses in Chapter 15, as identified in Table 15.0-7. The phrase to actuate ECCS within specified analytical limits is added at the requested location. The phrase process conditions is added, but with a slight rearrangement of the text for clarity. The changes are shown in the attached markup for FSAR Section 7.2.16.4. NuScale is also requested to add EQ testing and acceptance requirements to Part 8:
Part 8, License Condition; ITAAC, Section 2.4, Equipment Qualification - Module Specific For ITAAC No. 02.04.02, Add the same RPV riser level sensor performance requirements being proposed in the FSAR Section 7.2.16 (or add a reference to FSAR Sections 7.2.16 & 15.6.5, Tables 7.1-4 & 15.0-7) in Table 2.4-2 or Table 2.4-3 for RPV level elements, RCS-LE-1015A,B,C, & D Based on the revisions to the markup discussed above, FSAR Section 7.2.16.4 describes the performance requirements for the RPV riser level sensors. The results in FSAR Chapter 15, including sensitivity studies, show that the modeling of the RPV riser level sensors performance is not a key parameter. Therefore, additional requirements such as inspections, tests, analyses, and acceptance criteria (ITAAC) or combined license (COL) item are not warranted. Supplemental Response to Address NRC Clarification Call on November 26, 2024 As discussed during the NRC Clarification call, the RPV riser level sensors are located inside the RPV and are excluded from the Environmental Qualification Program. With respect to components inside the RPV, 10 CFR 50.49 (c)(3) states, environmental qualification of electric equipment important to safety located in a mild environment are not included within the scope of this section. A mild environment is an environment that would at no time be significantly more severe than the environment that would occur during normal plant operation, including anticipated operational occurrence. Consistent with regulation 10 CFR 50.49 (c)(3), as described in NuScale's response to audit item A-3.11-7, components inside the RPV are excluded from the Environmental Qualification Program. NuScale Nonproprietary NuScale Nonproprietary
Final Safety Analysis Report Section 3.1, Conformance with U.S. Nuclear Regulatory Commission General Design Criteria, identifies that the NuScale design conforms to General Design Criterion (GDC) 4 and identifies the following implementation in the NuScale Power Plant Design, the effects of environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents (LOCA), are considered in the design of safety-related and risk-significant SSC. The design protects against dynamic effects, including missiles, pipe whipping, and discharging fluids, that result from equipment failures and from events and conditions outside the NuScale Power Plant and prevents piping failure using break exclusion criteria. Consistent with GDC 4 conformance established in FSAR Section 3.1, components located inside the RPV comply with GDC 4. These components are designed to accommodate the effects of and to be compatible with the limiting environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including LOCAs. Specific to Instrumentation and controls (I&C), FSAR Section 7.1 identifies conformance with GDC 4. Therefore, it is not appropriate to add the RPV riser level sensor performance requirements to ITAAC 02.04.02. Though instrumentation inside the RPV is not subject to the Environmental Qualification Program, equipment qualification is still performed for equipment within the scope of GDC 4. As stated in FSAR Section 3.11, Environmental Qualification of Mechanical and Electrical Equipment, for electrical equipment located in mild environments, compliance with environmental design provisions of GDC 4 are achieved and demonstrated by proper incorporation of relevant environmental conditions into the design process, including the equipment specification. As stated in FSAR Section 7.2, System Features, the I&C structures, systems, and components are designed to perform their safety-related functional requirements over the range of environmental conditions postulated for the area that the components are located and during the time period when this performance is required. As appropriate, electrical equipment external to the RPV that is located in containment or other areas of the reactor building and meets the requirements of 10 CFR 50.49 (b)(1) and (b)(2) are subject to the Environmental Qualification Program. This equipment is identified in FSAR Table 3.11-1, List of Environmentally Qualified Equipment Located in Harsh Environments. NuScales Quality Assurance (QA) Program was inspected by the NRC from February 26 through March 1, 2024 to verify that NuScale effectively implemented an adequate QA program for design activities performed in support of the SDAA that comply with Appendix B to 10 CFR Part 50. The resulting inspection report (Adams No. ML24099A129) identifies the NRC staffs conclusion that: NuScale Nonproprietary NuScale Nonproprietary
NuScale is effectively implementing its QA program for activities affecting quality in support of the SDAA in accordance with the applicable requirements of Appendix B to 10 CFR Part 50 and 10 CFR Part 21. (Executive Summary)
NuScale is implementing its design control program in accordance with the regulatory requirements of Criterion III of Appendix B to 10 CFR Part 50. Based on the limited sample of documents reviewed, the NRC inspection team also determined that NuScale is adequately implementing their design control program in support of NuScales SDAA. (Section 2, Document Control)
NuScale has established its procurement document control and supplier oversight programs in accordance with the regulatory requirements of criterion IV and Criterion VII of Appendix B to 10 CFR Part 50. Based on the limited sample of documents reviewed, the NRC inspection team determined that NuScale is implementing its policies and procedures associated with the oversight of contracted activities in support of NuScales SDAA. (Section 6, Procurement Document Control and Control of Purchased Materials, Equipment and Components) The NRC safety evaluation for MN-122626-A, NuScale Power, LLC Quality Assurance Program Description, Revision 1, (Adams No. ML24033A318) made the safety finding that MN-122626-A adequately establishes controls that, when properly implemented comply with the requirements of Appendix B to 10 CFR Part 50. MN-122626-A, provides provisions for the following aspects:
As part of the overall design control process (Section 2.3, Design Control) a design verification process to ensure that items are suitable for their intended application and consistent with their effect on safety. Verification methods may include, but are not limited to, design reviews, alternative calculations, and qualification testing. Testing used to verify the acceptability of a specific design feature demonstrates acceptable performance under conditions that simulate the most adverse design conditions expected for the item's intended use. Design verification is completed before relying on the item to perform its intended design or safety function. (Section 2.3.1, Design Verification)
Procurement document control to ensure that purchased items, computer programs and services are subject to appropriate quality and technical requirements. Procurement document control includes, but is not limited to, Applicable technical, regulatory, administrative, quality, and reporting requirements (such as specifications, codes, standards, tests, inspections, special processes, and 10 CFR21) are invoked for procurement of items and services. 10 CFR 21 (or other applicable international equivalent) requirements for posting, evaluating, and reporting shall be followed and imposed on NuScale Nonproprietary NuScale Nonproprietary
suppliers when applicable. Applicable design bases and other requirements necessary to assure adequate quality shall be included or referenced in documents for procurement of items and services. (Section 2.4, Procurement Document Control)
Document control governing preparation of, issuance of, and changes to documents that specify quality requirements or prescribe how activities affecting quality are controlled to assure that correct documents are being employed. The types of documents that are controlled include, but are not limited to, purchase orders and related documents, vendor-supplied documents, inspection and test reports, technical specifications and nonconformance reports and corrective action reports. (Section 2.6, Document Control)
Purchased material, equipment and services control governing the acceptance of purchased items or services, such as source verification, receipt inspection, certificates of conformance, and document reviews (including Certified Material Test Report/Certificate). Acceptance actions and documents are established by the purchaser with appropriate input from the supplier and are completed to ensure that procurement, inspection, and test requirements, as applicable, have been satisfied before relying on the item to perform its intended safety function. (Section 2.7, Control of Purchased Material, Equipment, and Services)
Nonconforming materials, parts, or components controls governing items, including services, that do not conform to specified requirements to prevent inadvertent use. (Section 2.15, Nonconforming Materials, Parts, or Components) The above mentioned design verification process confirms the components ability to meet the design requirements. The QA program requires that the vendor provide a certificate of compliance to the procurement specification requirements, followed by NuScales documented review and acceptance. As discussed in the October supplemental response, the sensitivity cases confirm that LOCA analysis results are not sensitive to ECCS actuation modeling. Notwithstanding, the sensor performance and qualification of the sensor is verified through design control (e.g., design verification) included within NuScales robust Quality Assurance Program. Markups of the affected changes, as described in the response, are provided below. Note that while the associated document itself is identified as having export controlled information, none of the attached markup pages from the document contain export controlled information. NuScale Nonproprietary NuScale Nonproprietary
NuScale Final Safety Analysis Report Emergency Core Cooling System NuScale US460 SDAA 6.3-20 Draft Revision 1 Audit Question A-15.6.5-1 Table 6.3-1: Emergency Core Cooling System Actuation Values Parameter(2) Value(1) Low RPV Riser Level Table 7.1-4552 inches3 Low low Riser Level Table 7.1-4460 inches3 RPV low temperature & high pressure (LTOP) actuation The LTOP pressure setpoint is a function of the RCS cold temperature (Table 5.2-5 and Figure 5.2-3). Low AC Voltage timer Table 7.1-4 ECCS actuation delay after reactor trip Table 7.1-4 High high RCS pressure34 Table 7.1-4 High high RCS average temperature34 Table 7.1-4 Note 1: Additional information for ECCS actuation values is provided in Table 7.1-4. Note 2: Interlocks for these signals are described in FSAR Table 7.1-5. Note 3: Reference to the bottom of the pool. Note 34: ECCS actuation provides defense in depth.
NuScale Final Safety Analysis Report System Features NuScale US460 SDAA 7.2-72 Draft Revision 2 Ultrasonic flowmeter is used to measure RCS flow. Level Measurement Audit Question A-15.6.5-1 A guided wave radar method is used for measuring the PZR water level. The RPV riser level and CNV level measurements are thermal dispersion switch assemblies consisting of a probe that spans the distance of the measurement range with switches located at various levels to support operational requirements. The thermal dispersion switches determine the presence or absence of liquid water at discrete locations axially along the probe assembly. Modifications are made as needed to support qualification. The processing electronics for the radar unit is remote from the sensor assembly and located in a mild environment. 7.2.16.5 Installation Sensor signals in containment are transmitted through radiation and temperature tolerant mineral insulated cable and fed through designated containment electrical penetration assemblies. Outside the containment, the signals are routed to the processing electronics located in a mild environment. Power supplies are also housed in the processing cabinets in a mild environment. Reactor Coolant System Hot Temperature The sensors are installed in thermowells on the RPV vessel and inside containment. Four sensing elements, one per separation group, are mounted below the pressurizer baffle plate section to obtain safety-related RCS hot temperature measurement. Reactor Coolant System Cold Temperature Four RTDs, one per channel, are mounted in thermowells below the steam generators. The RTD lead wires are routed through a support conduit to the CNV penetration assemblies. Pressurizer Liquid Temperature The RTDs are mounted in thermowells in the lower section of the pressurizer. The RTD cabling is mineral insulated cable to protect the electronic signal from the containment environment conditions. The wiring is routed from the RTD in the thermowell location to the designated containment vessel electrical penetration assemblies and then to the MCS cabinets located in a mild environment outside containment. Pressurizer Vapor Temperature The two RTDs are installed in thermowells in the vapor section of the pressurizer. The RTD signals are protected from the containment environment by mineral insulated cables. The cablingwiring is routed from the RTD in the thermowell
NuScale Final Safety Analysis Report System Features NuScale US460 SDAA 7.2-74 Draft Revision 2 Main Steam Pressure Standard installation processes are used for these transmitters as the piping that the sensing lines tap into is outside of containment and above the reactor pool water level. Feedwater Pressure and Decay Heat Removal System Outlet Pressure The transmitters are located in the reactor pool and mounted to part of the DHRS condenser structure. This location allows them to be in close proximity to their respective sensing lines. The signal cable is waterproof cable that routes the signal from the transmitter to the electrical panel near the platform above the vessel. From there the signal goes to the MPS cabinets. Pressurizer Level The radar level sensor consists of an antenna that is the wave-guide for the transmitted signal. This assembly is inserted in the top of the reactor module for pressurizer level measurement such that its signal propagates down the antenna and reflects off the vapor/liquid surface. Reactor Pressure Vessel Riser Level Audit Question A-15.6.5-1 The sensors are mounted on top of the CNV with mineral insulated cable extending down to an instrument sealelectrical penetration assembly in the RPV head. A probe with multiple level switches extends downward through the PZR and into the RPV riser, spanning the distance of the measurement range, to monitor points below the baffle plate. Containment Water Level The level sensor consists of a single or series of level switches located at points determined by operational requirements. Decay Heat Removal System Level The sensor taps are fitted into the DHRS steam piping. The switch inserts into the pipe taps and sends the signal electronically to the disconnect panel using mineral insulated cable or cable of similar nature that is waterproof. From the disconnect panel the signals are sent to the MCS for MCR indication. Reactor Coolant System Flow The reactor coolant system flow sensors consist of ultrasonic flowmeter transducers mounted into a nozzle integral to the reactor vessel shell. These transducers are inserted into a well that is machined into the nozzle. This well is the reactor coolant pressure boundary and allows the transducer signal to pass through the well and perform its function.
NuScale Final Safety Analysis Report Fundamental Design Principles NuScale US460 SDAA 7.1-61 Draft Revision 2 Audit Question A-15.6.5.1 Table 7.1-4: Engineered Safety Feature Actuation System Functions ESFAS Function Process Variable Analytical Limit Number of Channels Logic Figure Showing the System Automated Function Emergency Core Cooling System (ECCS) Low ELVS voltage 24-hour Timer 24 hours 3 2/3 Figure 7.1-1n Low RPV Riser Level 540-552" in. (Note 4) 4 2/4 Low-Low RPV Riser Level 460-472 in. (Note 4) 4 2/4 ECCS Timer after Reactor Trip 8 hours 3 2/3 High-High RCS Average Temperature (Note 3) 620°F 4 2/4 High-High RCS Pressure (Note 3) 2500 psia 4 2/4 Decay Heat Removal System (DHRS) High Pressurizer Pressure 2100 psia 4 2/4 Figure 7.1-1l High Narrow Range RCS Hot Temperature 620°F 4 2/4 High Main Steam Pressure 1200 psia 4 2/4 Low AC Voltage to Battery Chargers 80% of normal ELVS voltage Actuation Delay of 60 seconds (Note 1) 4 2/4 High Narrow Range Containment Pressure 9.5 psia 4 2/4 Low Pressurizer Level 35% 4 2/4 High Under-the-Bioshield Temperature 250° F 4 2/4
NuScale Final Safety Analysis Report Fundamental Design Principles NuScale US460 SDAA 7.1-63 Draft Revision 2 Demineralized Water System Isolation Reactor Trip Any Reactor Trip Signal 4 2/4 Figure 7.1-1m Low RCS Flow 1.0 ft3/s 4 2/4 High Subcritical Multiplication (SCM) 3.2 4 2/4 Chemical and Volume Control System Isolation High Pressurizer Level 80% 4 2/4 Figure 7.1-1k Figure 7.1-1v Figure 7.1-1w High Narrow Range Containment Pressure 9.5 psia 4 2/4 Low Pressurizer Level 35% 4 2/4 Low-Low Pressurizer Level 15% 4 2/4 Low AC Voltage to Battery Chargers 80% of normal ELVS voltage Actuation Delay of 60 seconds (Note 1) 4 2/4 High Under-the-Bioshield Temperature 250°F 4 2/4 Pressurizer Heater Trip Low Pressurizer Level 35% 4 2/4 Figure 7.1-1m DHR actuation Any automatic DHR actuation 4 2/4 Low Temperature Overpressure Protection (LTOP) Low Temperature Interlock with High Pressure (WR RCS cold temperature and WR RCS Pressure) Variable based on WR RCS cold temperature and WR RCS Pressure as listed in Table 5.2-5 4 2/4 Figure 7.1-1n Pressurizer Line Isolation Low Pressurizer Pressure 1850 psia 4 2/4 Figure 7.1-1e, Figure 7.1-1h, Figure 7.1-1w Note 1: Normal AC voltage is monitored at the bus(es) supplying the battery chargers for the EDAS. Note 2: These signals provide automatic ECCS actuation for beyond-design-basis events. Note 3: These signals provide automatic ECCS actuation for beyond-design-basis events. The signals have non-safety related function; however, they are implemented using safety-related sensors. Note 4: Table 15.0-7 identifies how the RPV riser level analytical limits are modeled in the transient and accident analyses of Chapter 15. Table 7.1-4: Engineered Safety Feature Actuation System Functions (Continued) ESFAS Function Process Variable Analytical Limit Number of Channels Logic Figure Showing the System Automated Function
NuScale Final Safety Analysis Report System Features NuScale US460 SDAA 7.2-72 Draft Revision 2 Ultrasonic flowmeter is used to measure RCS flow. Level Measurement Audit Question A-15.6.5-1 A guided wave radar method is used for measuring the PZR water level. The RPV riser level and CNV level measurements are thermal dispersion switch assemblies consisting of a probe that spans the distance of the measurement range with switches located at various levels to support operational requirements. The thermal dispersion switches determine the presence or absence of liquid water at discrete locations axially along the probe assembly. Modifications are made as needed to support qualification. The processing electronics for the radar unit is remote from the sensor assembly and located in a mild environment. Audit Question A-15.6.5-1 The RPV riser level measurement provides input to MPS to actuate ECCS within the specified level range during service conditions typical of when ECCS actuation is expected. The RPV riser level instrumentation is qualified to actuate ECCS within specified analytical limits for the process conditions characteristic of the spectrum of postulated design-basis loss-of-coolant accidents (liquid and vapor) and the associated range of depressurizations. Additionally, qualification confirms the acceptability of the correlation between RPV riser level transient and the modeling approach used in the transient and accident analyses in Chapter 15, as identified in Table 15.0-7. 7.2.16.5 Installation Sensor signals in containment are transmitted through radiation and temperature tolerant mineral insulated cable and fed through designated containment electrical penetration assemblies. Outside the containment, the signals are routed to the processing electronics located in a mild environment. Power supplies are also housed in the processing cabinets in a mild environment. Reactor Coolant System Hot Temperature The sensors are installed in thermowells on the RPV vessel and inside containment. Four sensing elements, one per separation group, are mounted below the pressurizer baffle plate section to obtain safety-related RCS hot temperature measurement. Reactor Coolant System Cold Temperature Four RTDs, one per channel, are mounted in thermowells below the steam generators. The RTD lead wires are routed through a support conduit to the CNV penetration assemblies. Pressurizer Liquid Temperature
NuScale Final Safety Analysis Report System Features NuScale US460 SDAA 7.2-74 Draft Revision 2 Narrow Range Containment Pressure Four narrow range pressure transducers, one for each separation group, are installed in containment in four different locations, supplying four different separation groups. Wide Range Containment Pressure Two wide range pressure transducers are installed in containment in two locations, and cables are routed through separation groups B and C penetration assemblies. Main Steam Pressure Standard installation processes are used for these transmitters as the piping that the sensing lines tap into is outside of containment and above the reactor pool water level. Feedwater Pressure and Decay Heat Removal System Outlet Pressure The transmitters are located in the reactor pool and mounted to part of the DHRS condenser structure. This location allows them to be in close proximity to their respective sensing lines. The signal cable is waterproof cable that routes the signal from the transmitter to the electrical panel near the platform above the vessel. From there the signal goes to the MPS cabinets. Pressurizer Level The radar level sensor consists of an antenna that is the wave-guide for the transmitted signal. This assembly is inserted in the top of the reactor module for pressurizer level measurement such that its signal propagates down the antenna and reflects off the vapor/liquid surface. Reactor Pressure Vessel Riser Level Audit Question A-15.6.5-1 The sensors are mounted on top of the CNV with mineral insulated cable extending down to an instrument sealelectrical penetration assembly in the RPV head. A probe with multiple level switches extends downward through the PZR and into the RPV riser, spanning the distance of the measurement range, to monitor points below the baffle plate. The probe is routed through guide tubes within the RPV that contain openings to allow for vapor or liquid to be in contact with the probe. Containment Water Level The level sensor consists of a single or series of level switches located at points determined by operational requirements.
NuScale Final Safety Analysis Report Transient and Accident Analyses NuScale US460 SDAA 15.0-51 Draft Revision 2 Audit Question A-15.1.1-3, Audit Question A-15.6.5-1 Table 15.0-7: Analytical Limits and Time Delays Signal(1) Analytical Limit Basis and Event Type Actuation Delay High Power 115%(2) RTP ( 15% RTP) 25% RTP (<15% RTP) This signal is designed to protect against exceeding CHF limits for reactivity and overcooling events. 2.0 sec Source and Intermediate Range Log Power Rate 3 decades/min This signal is designed to protect against exceeding CHF and energy deposition limits during startup power excursions. Variable High Power Rate +/-7.5% RTP/30 sec This signal is designed to protect against exceeding CHF limits for reactivity and overcooling events. 2.0 sec High Source Range Count Rate 5.0 E+05 counts per second(3) This signal is designed to protect against exceeding CHF and energy deposition limits during rapid startup power excursions. 3.0 sec High Subcritical Multiplication 3.2 This signal is designed to detect and mitigate inadvertent subcritical boron dilutions in operating Modes 2 and 3. 150.0 sec High RCS Hot Temperature 620°F This signal is designed to protect against exceeding CHF limits for reactivity and heatup events. 8.0 sec High RCS Average Temperature 555°F This signal is designed to protect against exceeding CHF limits for reactivity events. 8.0 sec High Containment Pressure 9.5 psia This signal is designed to detect and mitigate RCS or secondary leaks above the allowable limits to protect RCS inventory and ECCS function during these events. 2.0 sec High Pressurizer Pressure 2100 psia This signal is designed to protect against exceeding RPV pressure limits for reactivity and heatup events. 2.0 sec High Pressurizer Level 80% This signal is designed to detect and mitigate CVCS malfunctions to protect against overfilling the pressurizer. 3.0 sec Low Pressurizer Pressure 1850 psia(4) This signal is designed to detect and mitigate high-energy line break (HELB) events from the pressurizer vapor space and protect RCS subcooled margin for protection against instability events. 2.0 sec Low-Low Pressurizer Pressure 1200 psia(5) This signal is designed to protect RCS subcooled margin for protection against instability events. 2.0 sec Low Pressurizer Level 35% This signal is designed to detect and mitigate pipe breaks to protect RCS inventory and ECCS functionality during LOCAs, primary HELB outside containment events, or SGTF, and to protect the pressurizer heaters from uncovering and overheating during decrease in RCS inventory events. 3.0 sec Low-Low Pressurizer Level 15% This signal is designed to detect and mitigate pipe breaks to protect RCS inventory and ECCS functionality during LOCAs, primary HELB outside containment events, or SGTF. 3.0 sec
NuScale Final Safety Analysis Report Transient and Accident Analyses NuScale US460 SDAA 15.0-53 Draft Revision 2 High Under-the-Bioshield Temperature 250°F This signal is designed to detect high energy leaks or breaks at the top of the NPM under the bioshield to reduce the consequences of HELBs on the safety-related equipment located on top of the module. 8.0 sec Notes:
- 1. Interlocks, permissives, and overrides for these signals are described in Table 7.1-5.
- 2. The overcooling event analyses account for decreased power detectionuncertainty due to decreasing downcomer temperature. The reductionuncertainty is 7%
for downcomer temperature decreases up to 10°F and is scaled upwards from 7% by 0.7%/ F for downcomer temperature decreases beyond 10°F.
- 3. The high count rate trip is treated as a source range over power trip that occurs at a core power analytical limit of 500 kW, which functionally equates neutron monitoring system counts per second to core power in watts. This trip is bypassed once the intermediate range signal is established.
- 4. If RCS hot temperature is above 500°F as shown in Figure 4.4-2.
- 5. If RCS hot temperature is below 500°F as shown in Figure 4.4-2.
- 6. If RCS hot temperature is above 500°F.
- 7. The RPV waterriser level is presented in terms of elevation where reference zero is the bottom of the module assembly (at the bottom of the reactor pool). The range accommodates instrumentation uncertainty. The RPV riser level is measured as described in Section 7.2.16. When riser level decreases to the setpoint within the analytical limit range, the actuation shall occur within the specified actuation delay time. In analyses where ECCS actuation occurs on RPV riser level, a low liquid void fraction in an analysis node corresponding to the specified actuation elevation is used as a surrogate for RPV riser level. Additional details of this modeling approach are provided in Section 15.6.5.
- 8. Normal AC voltage is monitored at the battery chargers for the EDAS. The analytical limit is based on an average voltage below 80% of normal.
Table 15.0-7: Analytical Limits and Time Delays (Continued) Signal(1) Analytical Limit Basis and Event Type Actuation Delay
NuScale Final Safety Analysis Report Decrease in Reactor Coolant Inventory NuScale US460 SDAA 15.6-14 Draft Revision 2 because DC power is still available, the MPS can still actuate these safety functions, including reactor trip, earlier if a separate actuation limit is reached. The event sequence for a loss of normal AC power is similar to that when no power is assumed lost. The primary difference is an earlier termination of secondary cooling. Thus, a loss of normal AC power conservatively maximizes the RCS thermal conditions after event initiation.
Loss of EDAS, normal DC power system (EDNS), and normal AC at event initiation - Power to the MPS is provided via the EDAS, so this scenario results in an immediate actuation of the reactor trip system, DHRS, SSI, containment isolation, opening of the RVVs, and opening of the RRVs when differential pressure between the RCS and the CNV drops below the IAB release pressure. Single failure evaluations of the failure of a single RVV to open, a failure of a single RRV to open, and failure of one ECCS division (one RVV and one RRV) to open were performed to determine the most conservative scenario. The evaluations show that the limiting single failure for some cases is the failure of one RVV and one RRV to open. The limiting MCHFR case is not affected by single failures because the minimum CHFR occurs before the single failures can occur. Allowances for instrument inaccuracy are accounted for in the analytical limits of mitigating systems in accordance with the guidance provided in RG 1.105. Audit Question A-15.6.5-1 The ECCS actuation on low RPV riser level is assumed to occur when liquid void fraction in a specific analysis node near the riser outlet decreases below 10 percent. This void fraction assumption, with the model nodalization, corresponds to ECCS actuation at a level of approximately 550 in., near the top of the analytical limit range in Table 15.0-7. Sensitivity cases are also included where ECCS actuation on low RPV riser level is assumed to occur when liquid void fraction in the next lower analysis node decreases below 10 percent. This void fraction assumption, with the model nodalization, corresponds to ECCS actuation at a level well below the bottom of the analytical limit range for low RPV riser level in Table 15.0-7. Audit Question A-15.6.5-1 The ECCS actuation on low-low RPV riser level is assumed to occur when liquid void fraction in a specific analysis node in the riser decreases below 5 percent. This void fraction assumption, with the model nodalization, corresponds to ECCS actuation at a level of approximately 473 in., near the top of the analytical limit range in Table 15.0-7. Additional sensitivity cases for modeling ECCS actuation at or below the bottom of the analytical limit range for low-low RPV riser level in Table 15.0-7 are not needed because the impact of ECCS actuation modeling on LOCA results is established based on the sensitivity cases for low RPV riser level described above. The input parameters and initial conditions used in the limiting case for the post-LOCA long-term cooling analysis are also selected to provide a
NuScale Final Safety Analysis Report Decrease in Reactor Coolant Inventory NuScale US460 SDAA 15.6-16 Draft Revision 2 Table 15.6-12 and shows the MCHFR values for each size break with and without AC and EDAS available. Other LOCA events have less limiting MCHFR values. Table 15.6-13 provides the limiting sensitivity parameters for each break type. In addition to MCHFR, the limiting sensitivities for collapsed liquid level (CLL), CNV pressure, and CNV temperature are provided. Note that these values consider loss of power, single failures, and the size of the break. However, the initial conditions are established to determine the lowest MCHFR. Section 6.2 contains the limiting CNV temperature and pressure analysis. The CLL is not challenged during the early phases of a LOCA. The limiting CLL is evaluated in the Extended Passive Cooling and Reactivity Control Methodology topical report (Reference 15.6-4) and the results are presented in Section 15.0.5. During the post-LOCA long-term cooling phase, the containment and RPV temperatures and pressures continue to decrease, indicating that the decay and residual heat are being removed from the RPV and containment. Audit Question A-15.6.5-1 The results of sensitivity cases show that ECCS actuation is delayed when the ECCS actuation is modeled below the bottom of the analytical range for low RPV riser level compared to when ECCS actuation is modeled near the top of the analytical range for low RPV riser level. For liquid space breaks, this delay does not typically prevent ECCS actuation from occurring on the low RPV riser level signal. For vapor space breaks, the additional cooldown from the break flow and DHRS during the delay typically causes the RCS cold temperature to decrease enough that the low RPV riser level signal is bypassed. If the low RPV riser level signal is bypassed, ECCS actuation is further delayed until the low-low RPV riser level is reached. Regardless of the signal that results in ECCS actuation, the additional cooldown results in a less limiting CNV response for sensitivity cases with delayed ECCS actuation. The MCHFR is not affected by the ECCS actuation modeling because MCHFR occurs before ECCS actuation on RPV riser level, either near event initiation or near the time of reactor trip. The sensitivity results show that CLL is not affected by the ECCS actuation modeling regardless of the timing or signal associated with ECCS actuation. The sensitivity cases confirm that LOCA results are not sensitive to ECCS actuation modeling. The MPS is credited to protect the NPM in the event of a LOCA. The following MPS signals provide the plant with protection during a LOCA: high pressurizer pressure high containment pressure low pressurizer level low pressurizer pressure low-low pressurizer pressure low-low pressurizer level
NuScale Final Safety Analysis Report Decrease in Reactor Coolant Inventory NuScale US460 SDAA 15.6-21 Draft Revision 2 system isolation, DHRS, and containment isolation. As power to the MPS is lost, both RVVs open immediately and the RRVs open when differential pressure between the RCS and the CNV drops below the IAB release pressure. The single failure evaluation considered one RVV failing to open, one RRV failing to open, or failure of one ECCS division causing one RVV and one RRV to fail to open. The evaluation compared the results to a scenario with no single failure. The evaluation showed that the single failure cases have no adverse impact on the limiting MCHFR. Therefore, the scenario with no single failure is limiting for the MCHFR analysis. The failure modes that could lead to a partial opening of an ECCS valve are characterized as having a remote probability of occurrence or are determined to not be credible. None of the credible component failure mechanisms that could prevent a full stroke of the ECCS valve have the potential to cause an ECCS valve to open. Therefore, a partial opening of an ECCS valve is not a credible initiating event. Allowances for instrument inaccuracy are accounted for in the analytical limits of mitigating systems in accordance with the guidance provided in RG 1.105. Audit Question A-15.6.5-1 The ECCS actuations on low RPV riser level and low-low RPV riser level are assumed to occur based on liquid void fraction as described in Section 15.6.5. Sensitivity cases with alternate ECCS actuation assumptions are also performed as described in Section 15.6.5. No operator action is credited. 15.6.6.3.3 Results Figure 15.6-42 to Figure 15.6-52 show the system response to an inadvertent ECCS operation event. Table 15.6-16 provides the results of the event. The limiting MCHFR case is initiated by an inadvertent opening of both RVVs. Upon the inadvertent opening of two RVVs, the large blowdown of steam into the containment causes rapid depressurization of the RCS and rapid pressurization of the containment. The RVV flow is shown in Figure 15.6-42. The increase in containment pressure initiates a reactor trip, containment isolation and secondary system isolation. The RPV and containment pressures are shown in Figure 15.6-43 and Figure 15.6-44, respectively. The rapid RCS depressurization causes voiding in the core and a decrease in RCS flow (Figure 15.6-45), leading to a reduction in CHFR (Figure 15.6-46). Reactor power decreases during this time due to control rod insertion and negative void feedback, as seen in Figure 15.6-47. Following the occurrence of transient MCHFR (Figure 15.6-46), a temporary increase in RCS flow is observed due to the increased density gradient from voiding in the riser (Figure 15.6-45).
NuScale Final Safety Analysis Report Decrease in Reactor Coolant Inventory NuScale US460 SDAA 15.6-22 Draft Revision 2 The isolation of the secondary system from high containment pressure causes an increase in steam generator pressure, as seen in Figure 15.6-48. DHRS actuates on high CNV pressure, however, DHRS does not actuate in time to have an impact on the MCHFR analysis. Heat transfer from the RCS to the secondary coolant isolated in the steam generator region is limited due to the decreasing RCS temperatures associated with decreasing pressure and saturation temperature. Steam generator pressure is not limiting for an inadvertent operation of ECCS event. As primary coolant is released to the containment through the open RVVs, the inventory level inside the containment increases. Pressure and temperature inside the RPV continue a gradual downward trend, as shown in Figure 15.6-43 and Figure 15.6-49. After the RRVs open and pressure equalizes across the RRVs, liquid coolant from the containment begins to flow into the RPV downcomer region. A two phase natural circulation loop is established through the ECCS valves with steam exiting the pressurizer area into containment through the RVVs and liquid returning from the containment to the RPV through the RRVs. Decay heat and residual heat is transferred from the containment to the reactor pool resulting in the pressure and the temperature inside the RPV and containment continuing to decrease. The transient continues until stable ECCS cooling is established with RCS pressure and temperature continuing to decrease. The module remains in a safe condition with liquid level maintained above the top of the core through the entire transient. The minimum collapsed liquid level (Figure 15.6-50) occurs after equilibrium conditions are established between the RPV and containment and is approximately 10 feet above the top of the active fuel. The fuel volume average temperature is shown in Figure 15.6-51, and the total reactivity for the event is shown in Figure 15.6-52. Audit Question A-15.6.5-1 The results of sensitivity cases show that ECCS actuation is delayed when the ECCS actuation is modeled below the bottom of the analytical range for low RPV riser level compared to when ECCS actuation is modeled near the top of the analytical range for low RPV riser level. For opening of an RRV, this delay does not typically prevent ECCS actuation from occurring on the low RPV riser level signal. For opening of one or two RVVs, remaining ECCS valves typically open on low differential pressure before ECCS actuation on low RPV riser level occurs. The sensitivity results show a less limiting CNV response with delayed ECCS actuation. The MCHFR is not affected by the ECCS actuation modeling because MCHFR occurs before ECCS actuation on RPV riser level, near event initiation or near the time of reactor trip. The sensitivity results show that CLL is not affected by the ECCS actuation modeling regardless of the timing or signal associated with ECCS actuation. The sensitivity cases confirm that inadvertent operation of ECCS results are not sensitive to ECCS actuation modeling.
Loss-of-Coolant Accident Evaluation Model TR-0516-49422-NP Draft Revision 4 © Copyright 2024 by NuScale Power, LLC 89 5.2 Analysis Setpoints and Trips A number of safety-related measurements exist in the NPM to detect off-normal conditions. Table 5-3 shows the typical safety-related measurements relevant to LOCA, IORV, and CNV pressure or temperature analysis along with their functions. The safety analysis analytical limits specify the setpoints, or range of setpoints, and the sensing delay for each safety-related signal. These parameters can be monitored in the NRELAP5 model to credit safety-related signal actuations. A single parameter may be used for multiple actuations. For example, a high containment pressure signal could result in reactor trip, containment isolation, and DHRS actuation. The MPS actuation signals and limits vary by NPM design and may be incorporated into the NRELAP5 model for the design. Audit Question A-15.6.5-1 (( }}2(a),(c) Table 5-3 Typical NuScale Power Module Safety-Related System Measurement Parameters (( }}2(a),(c)
Loss-of-Coolant Accident Evaluation Model TR-0516-49422-NP Draft Revision 4 © Copyright 2024 by NuScale Power, LLC 90 Audit Question A-15.6.5-1 The mixture level detection typically uses a simple approximation of the mixture level such as based on (( }}2(a),(c) Audit Question A-15.6.5-1 (( Audit Question A-15.6.5-1 Audit Question A-15.6.5-1 }}2(a),(c)
Loss-of-Coolant Accident Evaluation Model TR-0516-49422-NP Draft Revision 4 © Copyright 2024 by NuScale Power, LLC 91 (( Audit Question A-15.6.5-1 }}2(a),(c) 5.3 Initial Plant Conditions Table 5-4 provides the basis for conservatively biasing the initial conditions for LOCA analysis. These ranges are intended to account for both the normal control system deadband and the system/sensor measurement uncertainty without specifically quantifying the portion of the range applied to either uncertainty. For Phase 0 MCHFR analysis, the conservative bias directions are consistent, or a range of initial conditions are evaluated to confirm limiting bias directions. Table 5-4 Plant Initial Conditions (( }}2(a),(c)}}