ML24348A043
| ML24348A043 | |
| 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: ML24348A043 (1) | |
Text
Response to NuScale Topical Report Audit Question Question Number: A-NonLOCA.LTR-30 Receipt Date: 04/15/2024 Question:
(( 2(a),(c) Therefore, the applicant is expected to provide the bases for applicability of the EM for RCS pressurization and heat-up events, such as positive reactivity insertion and decrease in heat removal events. Comparisons of predicted and measured (or benchmarked) RCS pressure response for rapid pressurization events are needed. Provide the validation of the EM for RCS pressurization and heat-up events, such as positive reactivity insertion and decrease in heat removal events. For example, provide NRELAP5 code validation for a limiting RCS pressurization transient in well scaled test facilities. The code validations must show the similarities between the test facilities/apparatus and the reactor design. The results of the code validation must include the biases and uncertainties of the code in predicting the system phenomena on both the primary and secondary sides. The dependence of code biases on major parameters must be evaluated to ensure the code not to bias on the non-conservative direction. Sensitivity studies may be performed in lieu of tests. Provide proposed markups to the TR. NuScale Nonproprietary NuScale Nonproprietary
Response
As discussed in Section 5.3.7.2 of the non-loss-of-coolant accident (non-LOCA) topical report, TR-0516-49416-P Rev. 4, the purpose of the NIST-2 testing was to enhance the validation basis for the decay heat removal system (DHRS). The NIST-2 DHRS testing did include a loss of feedwater (LOFW) as part of the testing sequence and therefore has sometimes been referred to as an LOFW test. However, it is emphasized that the focus of the testing was on DHRS performance after operation of DHRS was underway. The specific initiating events and associated response prior to DHRS operation was not the focus of the testing. Therefore, comparison of the NIST-2 DHRS test response to the Final Safety Analysis Report (FSAR) Chapter 15 LOFW results is not appropriate. The NRELAP5 validation previously performed and included in the the non-LOCA topical report that was reviewed and approved by the NRC (TR-0516-49416-P-A Rev. 3) demonstrated applicability of the evaluation model to predict the reactor coolant system pressurization response during non-LOCA events. Justification is provided below. ((
}}2(a),(b),(c),ECI Contrast this to the sequence of events for the Chapter 15 LOFW in FSAR Table 15.2-17. In the Chapter 15 LOFW, the LOFW occurs at time zero but reactor trip does not occur until 13 seconds later. As a result, there is 13 seconds where the reactor is still generating full power NuScale Nonproprietary NuScale Nonproprietary
even though there is no feedwater being supplied to the steam generators. The mismatch between power being produced in the primary and power being removed by the secondary results in a significant heatup and pressurization of the primary during this period as shown in FSAR Figures 15.2-31 and 15.2-28, respectively. The pressurization results in the high pressurizer pressure analytical limit being reached and reactor trip occurs. Following reactor trip, power decreases rapidly to decay heat levels. The power reduction eliminates the mismatch between primary side power generation and secondary side heat removal, and pressure responds by decreasing rapidly as shown in FSAR Figure 15.2-28. After this rapid decrease, decay heat generation results in a smaller increase in pressure. The DHRS actuation valves fully open at 43 seconds per FSAR Table 15.2-17. The primary pressure is stable but somewhat elevated until effective DHRS cooling is established around 100 seconds as shown in FSAR Figure 15.2-29 and Figure 15.2-28. Once DHRS is in operation, pressure beings to slowly decrease as DHRS begins to effectively remove decay heat. ((
}}2(a),(c) This similarity during DHRS operation is expected as the intent of the testing was to focus on the period of DHRS operation as stated above.
Regarding the validation of NRELAP5 for pressurization and heatup events, the validation was previously performed and included in the the non-LOCA topical report that was reviewed and approved by the NRC (TR-0516-49416-P-A Rev. 3). ((
}}2(a),(c) As has been described in several previous audit responses, NuScale performed an applicability assessment and identified no new phenomena resulting from the design changes from the US600 design to the US460 design. Therefore, the existing validation basis (( }}2(a),(c) remains applicable. This audit question also requested information regarding code uncertainties associated with the above validation basis. As in the US600 design that was previously approved by the NRC, the maximum reactor coolant system pressure in the US460 design is limited by the reactor safety valves (RSVs) and there is little transient-dependent over-shoot as demonstrated by the limiting results presented in FSAR NuScale Nonproprietary NuScale Nonproprietary
Section 15.2. Therefore, code uncertainty has little impact on figures of merit associated with pressurization and heatup, provided the modeling of the RSV capacity and lift pressure are conservative. Because the validation results for pressurization and heatup events were previously reviewed by the NRC during approval of TR-0516-49416-P-A Rev. 3, are maintained in TR-0516-49416-P-A Rev. 4, and remain applicable, no new validation or markups are required. No changes to the SDAA are necessary. NuScale Nonproprietary NuScale Nonproprietary}}