ML24348A052
| ML24348A052 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 12/13/2024 |
| From: | NuScale |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24348A006 | List:
|
| References | |
| LO-176318 | |
| Download: ML24348A052 (1) | |
Text
Response to NuScale Topical Report Audit Question Question Number: A-NonLOCA.LTR-36 Receipt Date: 05/06/2024 Question:
The TR states: For transients where operation of the normal rod control function provides more adverse consequences of the transient, the control rod movement is modeled to increase the power response during the transient. However, the TR does not include methodology for achieving this goal. Please explain how to model the control rod movement, i.e., a methodology, to ensure that the modeling of rod movement results in more adverse consequences of the transient. For the same event, the staff notes that revision 3 of this TR states: For transients where operation of the normal rod control function provides more adverse consequences of the transient, the highest integral control rod bank worth (over time in cycle) is modeled, in conjunction with the maximum control bank withdrawal speed and bounding input for MTC, to increase the power response to the modeled control rod movement. The applicant needs to revise the TR to include the same level of specifications for the evaluation methods or new approach that includes sufficient conservatisms together with the technical bases.
Please provide proposed markups.
Response
The audit question identifies a change made in TR-0516-49416-P, Non-Loss-of-Coolant Accident Analysis Methodology, between Revision 3 (previously approved by the NRC) and Revision 4 (currently under review). Relevant excerpts from TR-0516-49416-P are identified below.
Section 4.3.1.1.1, Reactor Kinetics Model in NRC-approved Revision 3:
For transients where operation of the normal rod control function provides more adverse consequences of the transient, the highest integral control rod bank worth (over time in cycle) is modeled, in conjunction with the maximum control bank withdrawal speed and NuScale Nonproprietary NuScale Nonproprietary
bounding input for MTC, to increase the power response to the modeled control rod movement.
Section 4.3.1.1.1, Reactor Kinetics Model in Revision 4:
For transients where operation of the normal rod control function provides more adverse consequences of the transient, the control rod movement is modeled to increase the power response during the transient.
In the previous approach, maximum rod worth and maximum rod speed were both used to maximize the reactivity insertion associated with the control system response to temperature error (i.e., a temperature difference from the control setpoint). In the current approach, the maximum rod worth continues to be used. However, the rod speed is treated nominally. The use of a nominal rod speed, while a change to TR-0516-49416-P Section 4.3.1.1.1, is consistent with the discussion in TR-0516-49416-P Section 6.3.1.3 that states [t]he control rod controller uses design data to model a calculated rate of reactivity insertion due to maximum or nominal rod movement rates. The TR-0516-49416-P Section 6.3.1.3 discussion of using maximum or nominal rod movement rates in Revision 4 is unchanged from Revision 3 that was reviewed and approved by the NRC.
Final Safety Analysis Report (FSAR) Section 15.4.1.2 identifies that the maximum rod speed is 15 in/min and the nominal rod speed is 6 in/min. In the rod control system model, the rod speed increases as the absolute value of the temperature error increases, up to the nominal rod speed of 6 in/min for absolute value temperature errors of 5 degrees F and larger. ((2(a),(c) As a result of this modeling approach, the reactivity insertion caused by the rod control system (when enabled) varies with the size of the temperature decrease (i.e., cooldown). ((
}}2(a),(c)
Although the rod speed is limited to the nominal 6 in/min rather than the maximum 15 in/min, the overall approach is still conservative. The rod control design prohibits rod withdrawal at full NuScale Nonproprietary NuScale Nonproprietary
power as described in TR-0516-49416-P Sections 6.3.1.3 and 7.1.2. However, this design feature is ignored in the analyses where automatic rod control is assumed to exacerbate the transient. ((
}}2(a),(c) The result is that power can increase well above 100 percent due to the combination of the initiating event cooldown and the rod control response. For example, FSAR Figures 15.1-3, 15.1-12, 15.1-23, and 15.1-31 show a reactor power response above 100 percent power that is primarily driven by the rod control modeling with withdrawal allowed above 100 percent power.
Sections 4.3.1.1.1, 6.3.1.3, and 7.1.2 of TR-0516-49416-P are revised aMarkups 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. shown in the attached markups to provide details to ensure the conservatism of the modeling approach. 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 39
Effective delayed neutron decay constant of group i ( )
Prompt neutron generation time ( )
Reactivity feedback
Reactivity changes from control rod movement (normal controls), or scram
Decay heat model input In the non-LOCA evaluation model, the input to the point kinetics model is specified to give a conservatively high power response prior to actuation of reactor scram. The power response is biased by the input accounting for the reactivity feedback effects of moderator temperature, fuel temperature, and the normal control rod movement. After a reactor scram signal, the negative reactivity associated with insertion of the control and safety banks is conservatively modeled as described in Section 7.1.5. The moderator temperature coefficient (MTC) is a measure of the relative change in reactivity associated with a change in moderator (coolant) temperature. The Doppler temperature coefficient (DTC) is a measure of the relative change in the reactivity as the fuel temperature changes. In the non-LOCA evaluation model, reactivity feedback effects from moderator temperature changes and fuel temperature changes are conservatively bounded as described in Section 7.1.5. Negative reactivity insertion due to void generation impacts on moderator density is conservatively neglected. Audit Question A-NonLOCA.LTR-36 As described in Section 6.0 and Section 7.1.2, the rod control system and associated control logic are incorporated into the NRELAP5 model to allow simulation of reactivity changes (negative or positive) associated with normal control rod movement in response to postulated transients. In cooldown events, the normal rod control function attempts to increase average temperature by withdrawing control rods and therefore adding positive reactivity. For transients where operation of the normal rod control function provides more adverse consequences of the transient, the control rod movement is modeled to increase the power response during the transient. A conservative power response is ensured by using maximum rod worth and maximum rod speed for the control rod movement. A conservative power response is also ensured when using nominal rod speed if the 'insert only' mode is neglected as described in Section 6.3.1.3. The decay heat power contribution is conservatively bounded high or low, as appropriate for the specific transient, as described in Section 7.1.5. Appropriate input based on the core design is used for other parameters needed as input to the point reactor kinetics model. i
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 505 with a control rod controller and a boron controller. The control rod controller uses design data to model a calculated rate of reactivity insertion due to maximum or nominal rod movement rates. (( }}2(a),(c) Neither controller accounts for all the actual core physics including the effect of xenon or other decay products or poisons that could be expected with control rod repositioning. Audit Question A-NonLOCA.LTR-36 The average coolant temperature is controlled by adjusting core power, which is accomplished by moving the control rods or changing the boron concentration of the reactor coolant. The choice of which method is based on the desired rate of change for core power. The control rods are moved to achieve faster power changes to meet the target average coolant temperature; slower power changes are accomplished by changing the boron concentration of the reactor coolant. At full power, the rod control system is set to insert only mode to prevent automatic withdrawal of the control rods during a transient. If this design feature is neglected in the model, the rod control system will allow control rods to be withdrawn even if reactor power is at or above full power, resulting in a conservative power response to a transient. 6.3.1.4 Steam Pressure Control (Nonsafety-related) In an NPM design, the turbine throttle and bypass valves are used to control steam pressure at the programmed values, (( }}2(a),(c) 6.3.1.5 Feedwater and Turbine Load Control (Nonsafety-related) An NPM prototypic control scheme design for the feedwater system is based on turbine load demand. The feedwater pumps are variable speed and can provide variable flow for module operations over a wide range of power without adjustments to the feedwater regulating valve. (( }}2(a),(c)
Non-Loss-of-Coolant Accident Analysis Methodology TR-0516-49416-NP Draft Revision 5 © Copyright 2024 by NuScale Power, LLC 514 Core Average Coolant Temperature Control Audit Question A-NonLOCA.LTR-36 The average coolant temperature is controlled by moving the control rods or changing the boron concentration of the reactor coolant. This selection is based on the desired rate of change for core power. Hence, the control rods are moved to achieve faster power changes to meet the target average coolant temperature; slower power changes are accommodated by changing the boron concentration of the reactor coolant. At full power, the rod control system is set to insert only mode to prevent automatic withdrawal of the control rods during a transient. Neglecting this feature, if assumed, is conservative. Steam Pressure Control Steam pressure is controlled to the desired value using the turbine throttle valves or the turbine bypass valves. The effect of these valves to a change in steam pressure is considered for the non-LOCA transient analyses. Turbine Load Control The mass flow rate and pressure provided by the feedwater pump is used to meet the desired turbine load, which reflects the power generation rate. The impact of the feedwater pumps continuing to operate until the feedwater line is isolated is considered for the non-LOCA transient analyses when DHRS is actuated. Containment pressure control The containment pressure is established at sub-atmospheric conditions via operation of the containment evacuation system. The impact of this system continuing to operate is considered for the non-LOCA transient analyses. 7.1.3 Loss of Power Conditions This section defines the term loss of normal power as applied to an NPM; describes the various power supplies (AC and DC); and, explains how the loss of these power supplies is treated by the non-LOCA transient analyses. 7.1.3.1
Background
Chapter 15 of the SRP (Reference 15) does not ordinarily consider a loss of offsite power for events that require a malfunction of an active system for which power must be available; however, exceptions are made for some reactivity initiated events. The role of offsite power is less defined for an NPM plant than for traditional plants for several reasons, but the use of natural circulation for normal operation and safety systems is the fundamental reason. Consequently, these design features limit the impact of a power loss to an NPM plant compared to a traditional plant design that relies on forced circulation.}}