ML24348A040
| ML24348A040 | |
| 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: ML24348A040 (1) | |
Text
Response to NuScale Topical Report Audit Question Question Number: A-NonLOCA.LTR-28S Receipt Date: 09/12/2024 Question:
As a follow-up to audit question A-NonLOCA.LTR-28, NRC sent an advanced request for additional information (RAI) on the same topic on August 29, 2024. A clarification call to discuss the RAI was held on September 12, 2024. During the clarification call, NuScale took an action to provide a revised topical report markup to address the NRC concerns from the original audit question A-NonLOCA.LTR-28 and the advanced RAI.
Response
As described above, NuScale took an action to provide a markup to TR-0516-49416-P, Revision 4, Non-Loss-of-Coolant Accident Analysis Methodology, to address the NRC concerns from the original audit question A-NonLOCA.LTR-28 and the advanced request for additional information (RAI). The markup is attached to this response. A summary of the changes is provided for convenience as follows:
Section 4.3.5 is revised to replace the discussion of case-by-case with the explanation that the impact of pressure on critical heat flux is determined by varying the pressurizer pressure.
Table 7-7 (decrease in feedwater temperature) is revised to indicate that both the initial pressurizer pressure and pressurizer pressure control using spray are varied ((2(a),(c)
Table 7-14 (increase in feedwater flow) is revised to indicate that both the initial pressurizer pressure and pressurizer pressure control using spray are varied (( }}2(a),(c)
Table 7-19 (increase in steam flow) is revised to indicate that both the initial pressurizer pressure and pressurizer pressure control using spray are varied (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary
Table 7-24 (steam line break) is revised to indicate that both the initial pressurizer pressure and pressurizer pressure control using spray are varied (( }}2(a),(c) In addition, pressurizer pressure control using heaters is set to nominal (( }}2(a),(c) The initial pressurizer level is also revised to indicate that it is varied (( }}2(a),(c) Finally, an editorial error in the initial fuel temperature bias is corrected regarding the different fuel temperatures associated with beginning-of-cycle (BOC) and end-of-cycle (EOC).
Table 7-28 (containment flooding / loss of containment vacuum) is revised to indicate that pressurizer pressure control using spray is varied (( }}2(a),(c) In addition, pressurizer pressure control using heaters is set to nominal (( }}2(a),(c) Note that Table 7-28 already identified that initial pressurizer pressure is varied, so no change to that parameter is needed like it was in the other tables as described above.
Table 7-70 (dropped control rod assemblies) is revised to indicate that the initial pressurizer pressure is varied (( }}2(a),(c) Markups of the affected changes, as described in the response, are provided below: NuScale Nonproprietary NuScale Nonproprietary
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 43 These conditions are demonstrated for a reasonable time, typically a few hundred seconds, following the last safety system actuation expected to occur in the short term transient progression, to demonstrate that core cooling is established and conditions that result in minimum margin to the acceptance criteria have occurred. 4.3.5 Identification of Cases for Subchannel Analysis and Extraction of Boundary Condition Data For NPM non-LOCA events, VIPRE-01 is used for performance of subchannel analysis calculations. Reference 6, supplemented by Reference 28, describes the subchannel analysis methodology. In the transient subchannel analysis, the following acceptance criteria are assessed:
MCHFR
maximum fuel centerline temperature For non-LOCA events that require subchannel analysis, the NRELAP5 transient analyses provide the following input from the system transient calculation to the downstream subchannel analysis:
reactor power as a function of time
core exit pressure as a function of time
core inlet temperature as a function of time
total system flow rate as a function of time In the NRELAP5 system transient analysis, cases for downstream subchannel analysis are identified based on the conservative bias directions for the boundary condition input and considering an NPM natural circulation design. The conservative bias directions are discussed below, followed by a description of the methodology for identifying cases for downstream subchannel analysis. As identified in Reference 6, supplemented by Reference 28, for the system transient parameters provided by NRELAP5, the conservative bias directions to minimize the CHFR are as follows:
maximum reactor power (higher power increases the actual heat flux)
maximum core inlet temperature (higher temperature reduces energy addition needed to raise coolant to saturated conditions)
minimum system flow rate (minimum flow is conservative as there is less coolant flow in the reactor core available for heat transfer) Audit Question A-NonLOCA.LTR-28 The effect of pressure on critical heat flux (CHF) is established on a case by case basisby varying initial pressurizer pressure to determine the appropriate direction for biasing for events that are challenging for margin to CHF.
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 536 7.2.1.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms indicated in Table 7-7 are considered in identifying a bounding transient simulation for MCHFR.53 Audit Question A-NonLOCA.LTR-28 RAI Questions A-NonLOCA.LTR-31, A-NonLOCA.LTR-32, A-NonLOCA.LTR-46, A-NonLOCA.LTR-53, A-NonLOCA.LTR-56, A-NonLOCA.LTR-65 CHFR Due to the increase in reactor power and subsequent reduction of MCHFR, this acceptance criterion is challenged for the decrease in feedwater temperature event. Consequently, sensitivity cases are performed to support the follow-on MCHFR evaluation. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis, however the reactivity insertion rate from the cooldown event is insufficient to challenge the temperature limit. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. Table 7-7 Initial conditions, biases, and conservatisms - decrease in feedwater temperature Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied.Biased to the high condition }}2(a),(c) Table 7-6 Acceptance criteria - decrease in feedwater temperature (Continued) Acceptance Criteria Discussion
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 538 RCS Temperature Control Automatic rod control Varied. Boron concentration Not credited. PZR Pressure Control PZR spray Varied.Disabled. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Varied Turbine bypass valves Disabled. Table 7-7 Initial conditions, biases, and conservatisms - decrease in feedwater temperature (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 546 7.2.2.3 Biases, Conservatisms, and Sensitivity Studies Audit Question A-NonLOCA.LTR-60 The biases and conservatisms indicated in Table 7-14 are considered in identifying a bounding transient simulation for criteria other than SG overfill. Audit Questions A-NonLOCA.LTR-28, A-NonLOCA.LTR-65 RAI Questions, A-NonLOCA.LTR-31, A-NonLOCA.LTR-32, A-NonLOCA.LTR-46, A-NonLOCA.LTR-53, A-NonLOCA.LTR-56, A-NonLOCA.LTR-65 Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. SG overfill An increase in feedwater flow can lead to overfilling the SG. If the SG is overfilled, DHRS performance is negatively impacted. This criterion is satisfied by demonstrating the SG does not overfill and DHRS cooling is adequate, consistent with the discussion in the acceptance criterion for escalation of an AOO to an accident. Table 7-14 Initial conditions, biases, and conservatisms - increase in feedwater flow Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied.Biased to the high condition. Initial PZR level Biased to the high condition. }}2(a),(c) Table 7-13 Acceptance criteria - increase in feedwater flow (Continued) Acceptance Criteria Discussion
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 548 RCS Temperature Control Automatic rod control Enabled. Boron concentration Not credited. PZR Pressure Control PZR spray Varied.Disabled. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled to control to constant steam pressure. Turbine bypass valves Disabled. Feedwater and Turbine Load Control Table 7-14 Initial conditions, biases, and conservatisms - increase in feedwater flow (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 558 Audit Question A-NonLOCA.LTR-28 RAI Questions A-NonLOCA.LTR-31, A-NonLOCA.LTR-32, A-NonLOCA.LTR-46, A-NonLOCA.LTR-53, A-NonLOCA.LTR-56, A-NonLOCA.LTR-65 Table 7-19 Initial conditions, biases, and conservatisms - increase in steam flow Parameter Bias / Conservatism Basis (( Initial reactor power Varied. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied.Biased to the high condition. Initial PZR level Biased to the high condition. Initial feedwater temperature Nominal. Initial fuel temperature Nominal. MTC Biased to EOC conditions. Kinetics Biased to the EOC condition. Decay heat Biased to the high condition. Initial SG pressure(1) Biased to the high condition. SG heat transfer Nominal RSV lift setpoint Nominal SG tube plugging Biased to the low condition. Steam flow increase Varied }}2(a),(c)
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 559 RCS Temperature Control Automatic rod control Varied. Boron concentration Not credited. PZR Pressure Control PZR spray Varied.Disabled. PZR heaters Nominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves N/A. Turbine bypass valves N/A. Feedwater and Turbine Load Control feedwater pump speed Disabled. Table 7-19 Initial conditions, biases, and conservatisms - increase in steam flow (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 564 7.2.4.3 Biases, Conservatisms, and Sensitivity Studies RAI Questions A-NonLOCA.LTR-31, A-NonLOCA.LTR-32, A-NonLOCA.LTR-46, A-NonLOCA.LTR-53, A-NonLOCA.LTR-56, A-NonLOCA.LTR-65 The biases and conservatisms indicated in Table 7-24, and a spectrum of break locations and sizes, are considered in identifying a bounding transient simulation for MCHFR and mass release. In the radiological analysis, primary-to-secondary leakage is assumed, which allows primary coolant to be released with the SG break flow. The transient SG mass release can be calculated for use as an input to the downstream radiological analysis. Alternatively, bounding assumptions for primary coolant release can be used in the radiological analysis to eliminate the need for calculating SG mass release. If such bounding assumptions are used, transient analysis to maximize SG mass release is not required. Audit Question A-NonLOCA.LTR-28 Secondary pressure Due to the depressurizing nature of this cooldown event, secondary pressure remains below the acceptance criterion for peak secondary pressure. Critical heat flux ratio This criterion is evaluated by downstream subchannel analysis. Sensitivity cases are performed to support the follow-on MCHFR evaluation. Maximum fuel centerline temperature This criterion is evaluated by downstream subchannel analysis, however the reactivity insertion rate from the cooldown event is insufficient to challenge the temperature limit. Containment integrity Containment integrity is evaluated by a separate analysis methodology. Escalation to a more serious accident or consequential loss of functionality This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. Table 7-24 Initial conditions, biases, and conservatisms - steam line break Parameter Bias / Conservatism Basis(1) (( Initial reactor power Biased upwards to account for measurement uncertainty. }}2(a),(c) Table 7-23 Acceptance criteria - steam line break (Continued) Acceptance Criteria Discussion
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 565 Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied.Biased to the high condition. Initial PZR level Varied.Biased to the high condition. Initial feedwater temperature Varied. Initial fuel temperature Biased low for BEOC conditions and high for EBOC conditions. MTC Both EOC and BOC conditions. Kinetics Both EOC and BOC conditions. Decay heat Biased to the high condition. Initial SG pressure(2) Varied. SG heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition. RCS Temperature Control Automatic rod control Enabled. (MCHFR) Boron concentration Not credited. Table 7-24 Initial conditions, biases, and conservatisms - steam line break (Continued) Parameter Bias / Conservatism Basis(1) (( }}2(a),(c)
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 566 PZR Pressure Control PZR spray Varied.Disabled. PZR heaters EnabledNominal. PZR Level Control Charging Not credited. Letdown Disabled. Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Disabled. Table 7-24 Initial conditions, biases, and conservatisms - steam line break (Continued) Parameter Bias / Conservatism Basis(1) (( }}2(a),(c)
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 571 7.2.5.3 Biases, Conservatisms, and Sensitivity Studies The biases and conservatisms presented in Table 7-28 are considered in order to identify the bounding transient simulation for MCHFR for the CNV flooding/loss of CNV vacuum event. Audit Question A-NonLOCA.LTR-28 RAI Questions A-NonLOCA.LTR-31, A-NonLOCA.LTR-32, A-NonLOCA.LTR-46, A-NonLOCA.LTR-53, A-NonLOCA.LTR-56, A-NonLOCA.LTR-65 Escalation of an AOO to an accident This criterion is satisfied by demonstrating stable RCS flow rates and constant or downward trending RCS and DHRS pressures and temperatures exist at the end of the transient, all acceptance criteria evaluated in the transient analysis are met, and shutdown margin is maintained at the end of the transient. RCS conditions during extended DHRS cooling are addressed in a separate analysis. Table 7-28 Initial conditions, biases, and conservatisms - containment flooding / loss of containment vacuum Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied Initial PZR level Nominal Initial feedwater temperature Nominal Initial fuel temperature Nominal MTC Biased to the EOC condition. Kinetics Biased to the EOC condition. Decay heat Biased to the high condition. Initial SG pressure(1) Nominal SG heat transfer Nominal }}2(a),(c) Table 7-27 Acceptance criteria - containment flooding / loss of containment vacuum Acceptance Criteria Discussion
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 572 RSV lift setpoint Nominal SG tube plugging Biased to the low condition. Initial containment pressure Varied Initial pool temperature Varied RCCW leak flow Varied RCCW temperature Varied RCS Temperature Control Automatic rod control Enabled. Boron concentration Not credited. PZR Pressure Control PZR spray Varied.Disabled. PZR heaters EnabledNominal. Table 7-28 Initial conditions, biases, and conservatisms - containment flooding / loss of containment vacuum (Continued) Parameter Bias / Conservatism Basis (( }}2(a),(c)
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 633 Audit Question A-NonLOCA.LTR-28 Table 7-70 Initial conditions, biases, and conservatisms - control rod misoperation, dropped control rod assemblies Parameter Bias / Conservatism Basis (( Initial reactor power RTP biased upwards to account for measurement uncertainty. Initial RCS average temperature Biased to the high condition. Initial RCS flow rate Biased to the low condition. Initial PZR pressure Varied.Biased to the high condition. Initial PZR level Nominal. Initial feedwater temperature Nominal. Initial fuel temperature Nominal. MTC Varied. Kinetics Varied. Decay heat Biased to the high condition. Initial SG pressure Nominal. SG heat transfer Nominal. RSV lift setpoint Biased to the high condition. SG tube plugging Biased to the low condition. Dropped CRA worth Minimum }}2(a),(c)}}