ML24348A018
| ML24348A018 | |
| 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: ML24348A018 (1) | |
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Response to NuScale Topical Report Audit Question Question Number: A-NonLOCA.LTR-5S2 Receipt Date: 04/29/2024 Question:
Audit question A-NonLOCA.LTR-5 was received by NuScale on April 24, 2023. NuScale provided a response on May 3, 2023. On February 23, 2024, NuScale received written feedback from the NRC. NuScale provided a supplemental response on February 27, 2024 to address the feedback. On April 29, 2024, NuScale received the following additional written feedback from the NRC: (( 2(a),(c) NuScale Nonproprietary NuScale Nonproprietary
(( }} 2(a),(c)
Response
This response is a supplement to both the May 3, 2023 original response and the February 27, 2024 supplemental response; the May 3, 2023 and February 27, 2024 responses remain unchanged in the electronic reading room (eRR). To address the detailed NRC second round of feedback, each aspect of the feedback is addressed by providing the text of the feedback in indented, italicized text, followed by the NuScale response in un-indented, regular text. (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary
((
}}2(a),(c) Using SIMULATE5 is appropriate for these static calculations and consistent with Section 4.3.1.1 of TR-0516-49416, Revision 4, Non-Loss-of-Coolant Accident Analysis Methodology, that states Core design analysis performed in accordance with a methodology approved for a NuScale design provides input to the system transient analysis. The rod drop methodology discussion in Section 7.2.15.1 of TR-0516-49416 is revised as shown in the attached markup to reference Section 4.3.1.1 to use an approved core design method and describe how to calculate the parameters for the rod drop.
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}}2(a),(c)
If rod drop is bounded by single rod withdrawal, then there is no specific transient calculation using NRELAP5 and VIPRE-01 for the rod drop event. If rod drop is not bounded, then the previously approved rod drop methodology in the topical report is used; that methodology does include NRELAP5 and VIPRE-01 transient calculations. For the Standard Design Approval Application (SDAA), the rod drop cases are confirmed to be bounded by single rod withdrawal and therefore Final Safety Analysis Report (FSAR) Section 15.4.3.5.1 identifies that no specific transient analysis with NRELAP5 and VIPRE-01 is needed for rod drop. However, FSAR Section 15.4.3.5.1 and Section 15.4.3.5.2 are revised as shown in the attached markup to clarify NuScale Nonproprietary NuScale Nonproprietary
that the topical report only provides the method for how to determine if rod drop is bounded by other events; the actual confirmation is performed in the analysis that implements the methodology. In this case, the analysis results confirm the rod drop is bounded as indicated in FSAR Section 15.4.3.5.3. ((
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(( }}2(a),(c) Note that the discussion in Section 7.2.15.1 of TR-0516-49416 does not make a generic conclusion about the size of the power overshoot. Instead, Section 7.2.15.1 of TR-0516-49416 identifies that comparison to the single rod withdrawal analysis is made to account for the possibility of power overshoot. Therefore, TR-0516-49416 does not require further revision or clarification. ((
}}2(a),(c)
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((
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Note that the feedback also identified that certain previous descriptions should be docketed or incorporated into the TR. The audit has an existing process where the NRC can request that a response to an audit question be docketed. If the NRC believes that information in a NuScale response should be docketed, the NRC should use the existing process to make that request. (( NuScale Nonproprietary
}}2(a),(c)
Markups of the affected changes, as described in the response, are provided below:
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 598 Sensitivity studies are performed as needed to identify the limiting response(s) for the acceptance criteria parameter(s) challenged by the event. Consequently, a sensitivity study is generally performed to identify cases for lowest MCHFR for this reactivity event. 7.2.15 Control Rod Misoperation The methodology used to simulate a postulated control rod misoperation for an NPM, and an evaluation of the acceptance criteria for an AOO listed in Table 7-4, are presented below. 7.2.15.1 General Event Description and Methodology The rod control system is used to move (insert or withdraw) the control rod assemblies (CRAs) in response to an operator action or an automatic control. Since these transients are initiated by a malfunction in the rod control system, a Steam Pressure Control Turbine throttle valves Enabled. Turbine bypass valves Disabled. Feedwater and Turbine Load Control feedwater pump speed Enabled. CNV Pressure Control CNV evacuation system Enabled.
- 1. ((
}}2(a),(c) Table 7-65 Not Used Table 7-64 Initial conditions, biases, and conservatisms - uncontrolled control rod bank withdrawal at power (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 599 variety of reactivity related conditions can result. Specific reactivity conditions for an NPM include: 1) withdrawing a single CRA; 2) dropping one or more CRAs; or,
- 3) leaving one or more CRAs behind when inserting or withdrawing a control bank. The consequences for each of these reactivity conditions are discussed below.
Table 7-66 lists the relevant acceptance criteria, single active failure, and loss of power scenarios. Withdrawal of a Single CRA The withdrawal of a single CRA causes a reactivity insertion that increases reactor power and leads to a rise in coolant temperature, pressurizer level, and RCS pressure. Feedback from the rising fuel temperature is not sufficient to counteract the reactivity insertion, so the power increases until the system trips on high power, high power rate, high pressurizer pressure, or high RCS temperatures. The maximum power and minimum MCHFR occur just after the resulting scram, while the peak primary pressure occurs some time later, as the DHRS begins to function and remove heat through the steam generators. Finally, stable DHRS cooling is established at the end of the transient. The limiting MCHFR typically occurs for a reactivity insertion rate that results in reactor trip on core power, pressurizer pressure, and RCS temperature signals at approximately the same time. These conditions arise for events initiated from partial power with lower reactivity insertion rates because higher reactivity insertion rates cause the MPS to trip much earlier on high power rate. The earlier reactor trip reduces the energy added to the reactor coolant, thereby producing a higher MCHFR. The asymmetry associated with the core power response causes the ex-core detectors to respond differently for each quadrant. Consequently, the range of reactivity insertion rates considered is sufficient to identify the point of transition to the high power rate signal (using the lowest reading ex-core detector based on the minimum after to before event initiation ratio of the radial peaking factors for the outer row of fuel assemblies). Sensitivity studies are performed on a variety of parameters, as necessary, to identify the case(s) with a potentially limiting MCHFR. The NRELAP5 MCHFR pre-screening process is employed to identify the cases sent for a detailed subchannel evaluation. In contrast to the uncontrolled CRA bank withdrawal at power event (Section 7.2.14), the highest reactivity insertion rate as determined from the resulting rod worth and control rod step speed is significantly lower for the withdrawal of a single CRA event. Audit Question A-NonLOCA.LTR-5 Note that instead of using case-specific values for rod worth and parameters associated with the asymmetry, more conservative bounding input values can be used. The use of bounding input values provides a conservative analysis simplification for the withdrawal of a single CRA event. In addition, the bounding
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 600 input values can then be used to compare this event to the dropping of one or more CRAs event described below. Dropping One or More CRAs Based on the minimum worth at any time during the cycle for a given core power, i.e., with the control bank positioned at the PDIL, dropping a single CRA causes a reactivity insertion that decreases reactor power. Feedback from the decreasing fuel temperature and the actions of the rod control system to restore power are generally not sufficient to counteract the reactivity insertion, so the power decreases until the system trips on high power rate. For event scenarios without a return to power, the maximum core power, peak primary pressure, and MCHFR occur at event initiation. The peak secondary system pressure occurs some time after the scram, as the DHRS begins to function and remove heat through the steam generators. Finally, stable DHRS cooling is established at the end of the transient. The potential for a return to power exists only for events initiated from less than RTP because the reduced worth of the dropped rod gives the rod control system time to act. The corresponding MCHFR for a dropped rod event with a return to power is typically greater than the MCHFR for events initiated from HFP. Hence, the limiting MCHFR cases typically occur at HFP conditions. Following reactor trip and subsequent turbine trip, the turbine bypass to the condenser opens to control the RCS temperature. However, the actions of the turbine bypass system are not credited, so as to minimize heat removal by the secondary side. Although turbine load is an input to the feedwater controller, no changes are made to this controller because the RCS responses are not sufficient to affect feedwater control. The asymmetry associated with the core power response causes the ex-core detectors to respond differently for each quadrant. The power input to the high power rate signal uses the highest reading ex-core detector, multiplying the core average power by the maximum after drop to before drop ratio of the radial peaking factors for the outer row of fuel assemblies. Sensitivity studies are performed on a variety of parameters, as necessary, to identify the case(s) with a potentially limiting MCHFR. The NRELAP5 MCHFR pre-screening process is employed to identify the cases sent for a detailed subchannel evaluation. The maximum fuel centerline temperature typically occurs at event initiation for those event scenarios with an immediate reactor trip. If the event scenario has a return to power, the maximum fuel centerline temperature is typically bounded by the fuel centerline temperature at HFP because the associated power peak is less than full power. Audit Question A-NonLOCA.LTR-5 As an alternative to performing a system transient analysis, the MCHFR and linear heat generation rate of the dropped rod event can be confirmed to be
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 601 bounded by other events. Most rod drops result in a reactor trip on high power rate because of the immediate decrease in power from the dropped rod. Figure 7-3 shows the decrease in power from the dropped rod (( }}2(a),(c) as a function of rod worth for a variety of initial power levels. The figure is based on a representative NPM core design and also overlays a representative high power rate trip. (( }}2(a),(c) Sensitivity studies were performed to confirm this behavior using several representative NPM core designs, a range of initial power levels, and a variety of rod worths. (( }}2(a),(c) The alternative bounding method for the rod drop uses this result to screens rod drop cases into two groups: those that result in reactor trip within a short period and those that do not. Audit Question A-NonLOCA.LTR-5 For the larger group of cases that result in the early reactor trip (( }}2(a),(c), the dropped rod, and subsequent reactor trip, cause a decrease in global power. The dropped rod does cause an asymmetry and results in an increase in local power peaking (primarily a radial increase with a small axial increase). The subsequent reactor trip eliminates the asymmetry and the associated local power peaking. Figure 7-4 shows the impact on the local power from the decrease in global power and the increase in local peaking for a representative NPM rod drop case that results in early reactor trip. (( }}2(a),(c) The relevant acceptance criteria in Table 7-66 for the first group of the rod drop event can be assessed by the limiting steady-state subchannel analysis. Audit Question A-NonLOCA.LTR-5 For the smaller group of cases that do not result in the early reactor trip (( }}2(a),(c) However, because power overshoot may occur during the event, these cases are bounded instead by comparison to the withdrawal of a single CRA. (( A core design code consistent with the discussion in Section 4.3.1.1, such as SIMULATE5 (Reference 23), is used to calculate (( }}2(a),(c)
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 602 (( }}2(a),(c) the conditions associated with the dropped rod cases are bounded by the valueslimits used in the analysis of the withdrawal of a single CRA. Therefore, the limiting MCHFR and linear heat generation rate for the dropped rod cases that do not result in early reactor trip are bounded by the results of the single rod withdrawal analysis. This comparison of the rod drop analysis cases to the single rod withdrawal analysis values is performed on a designcycle-specific basis. A rod drop analysis case with a parameterpoint that falls outside the values used in the single rod withdrawal analysis limits, (( }}2(a),(c), would require use of the system transient approach described earlier in this section. When the designcycle-specific comparison is bounding, the relevant acceptance criteria in Table 7-66 for the second group of the rod drop event can be assessed by the limiting single rod withdrawal subchannel analysis. Misalignment of One or More CRAs The misalignment of CRAs occurs as a result of one or more CRAs being left behind when inserting or withdrawing the control bank. These conditions are not evaluated with NRELAP5 as part of the non-LOCA event methodology because this event is not a transient. Instead, the MCHFR is determined as part of a detailed subchannel evaluation.
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 605 Audit Question A-NonLOCA.LTR-5 Figure 7-5 Example comparison of parameters associated with non-tripped rod drop events cases to single rod withdrawal event analysis valueslimits (( }}2(a),(c)
NuScale Final Safety Analysis Report Reactivity and Power Distribution Anomalies NuScale US460 SDAA 15.4-13 Draft Revision 2 from the rising fuel and moderator temperatures partially counteracts the reactivity insertion, slowing the power increase. For CRM cases with higher reactivity insertion rates, the MPS trips the reactor on high power or high power rate. These cases are non-limiting because the reactor is tripped before the maximum amount of reactivity can be inserted. The limiting combination of reactivity insertion and reactivity feedback produces the maximum possible power increase without reaching the high power or high power rate limits. The power increase is terminated by MPS actuation of reactor trip when the high RCS hot temperature limit, high pressurizer pressure limit, or high RCS average temperature limit is reached. The MPS high pressurizer pressure signal trips the reactor and actuates the DHRS for the limiting MCHFR case. The most limiting MCHFR (Figure 15.4-16) occurs at the moment before the power begins to decrease. The MCHFR remains above the CHF analysis limit and no fuel centerline melting is predicted for the withdrawal of a single CRA. The LHGR calculated for the single CRA withdrawal is below the calculated limits for cladding strain and fuel centerline melting. The maximum RCS pressure occurs after MCHFR occurs and is followed by decreasing RCS temperature and pressure. The limiting LHGR occurs in a single CRA withdrawal case where the initial power and limiting reactivity insertion rate is the same as that for the MCHFR case, and does not cause a high power rate trip like cases at lower initial powers. The pressurizer spray is assumed to function normally, which delays the trip on high pressure. This case maximizes power in combination with asymmetric peaking, resulting in a limiting LHGR. The LHGR remains below the design limit, so no fuel centerline melting is predicted for the single CRA withdrawal. 15.4.3.5 Control Rod Assembly Drop Thermal Hydraulic and Subchannel Analysis 15.4.3.5.1 Evaluation Models Audit Question A-NonLOCA.LTR-5 There is no NPM thermal hydraulic analysis required for the CRA drop if it is determined to be bounded. The Non-LOCA Analysis Methodology Topical Report (Reference 15.4-3) identifies thatprovides the methodology to determine if the CRA drop is bounded by either steady-state conditions at event initiation or by the single CRA withdrawal in Section 15.4.3.4. SIMULATE5 is used to calculate CRA worth and power distributions. A discussion of SIMULATE5 is provided in Section 4.3. The subchannel core CHF analysis of the events bounding the CRA drop are performed using VIPRE-01. VIPRE-01 is a subchannel analysis tool designed for general-purpose thermal-hydraulic analysis under normal operating conditions, operational transients, and events of moderate severity. Section 15.0.2 includes a discussion of the VIPRE-01 code and evaluation model.
NuScale Final Safety Analysis Report Reactivity and Power Distribution Anomalies NuScale US460 SDAA 15.4-14 Draft Revision 2 15.4.3.5.2 Input Parameters and Initial Conditions Audit Question A-NonLOCA.LTR-5 A spectrum of possible CRA drop scenarios is analyzed to determine ifensure the scenarios are bounded by either steady-state conditions at event initiation or by the single CRA withdrawal in Section 15.4.3.4. The following initial conditions and assumptions ensure the results have sufficient conservatism. The initial power levels of 20 percent, 50 percent, 75 percent, and 100 percent of nominal power are analyzed. Scenarios are analyzed for drop of a single CRA within a group and for drop of a whole group. The BOC, MOC, and EOC core conditions are analyzed. Other initial conditions varied in the analyzed cases include initial CRA position, axial offset, flow, and temperature. For the initial CRA drop, a representative drop time is used for all cases. For cases where a reactor trip occurs, conservative reactor trip characteristics are used, including a maximum time delay and holding the most reactive rod out of the core. Allowances for instrument inaccuracy are accounted for in the analytical limits of mitigating systems in accordance with RG 1.105. 15.4.3.5.3 Results Following a CRA drop, there is a rapid decrease in core reactivity and power. The high power rate limit is reached in the first few seconds of the transient. The MPS sends a reactor trip signal, terminating the event. At lower powers, the power decrease is less pronounced, and the reactor does not trip. In these cases, the regulating CRA bank brings the reactor back to the initial power after an initial power overshoot. Of the analyzed cases, most are confirmed to reach the high power rate limit by 1.5 seconds. Although local peaking increases due to the asymmetry caused by the CRA drop, total power decreases throughout the event due to the initial CRA drop and subsequent reactor trip. For these cases, the decrease in total power is greater than the increase in local peaking at all times such that local power decreases throughout the event. With the transient decrease in local power, the MCHFR and LHGR are limiting at the initial conditions and therefore bounded by the limiting steady-state subchannel analysis. The cases that do not reach the high power rate limit by 1.5 seconds are characterized by low initial power and low CRA worth. Because of the lower initial power, these cases are also likely bounded by the limiting steady-state subchannel analysis. However, because power overshoot may occur during the event, these cases are instead compared to the single CRA withdrawal event. The single CRA withdrawal limits for CRA worth, radial peaking, and}}