ML24346A245
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Response to SDAA Audit Question Question Number: A-15-2 Receipt Date: 09/05/2023 Question:
SDAA Subsection 7.0.4.5 contains COL Item 7.0-1. Col Item 7.0-1 states:
An applicant that references the NuScale Power Plant US460 standard design will demonstrate the stability of the NuScale Power Module during normal and power maneuvering operations for closed-loop module control system subsystems that use reactor power as a control input.
COL Item 7.0-1 indicates that a COL applicant will demonstrate the reactor power response is stable for the NPM at all normal and power maneuvering operations from 0 to 100 percent power using secondary side controllers. However, Section 5.4.1.3 of the SDAA states:
Analyses regarding the susceptibility of the NPM to develop DWO conditions use the approach documented in Appendix B of TR-131981-P, Methodology for the Determination of the Onset of Density Wave Oscillations (DWO), Reference 5.4-11. Results show that the combination of operating conditions and inlet flow restrictor design allow for margin to DWO onset at all nominal power levels from 20 percent to 100 percent power, which is the power generation range for turbine operation. While DWO may occur during limited operational times at low power levels, the SG and inlet flow restrictor design assures that DWO transient conditions are acceptable to meet applicable ASME BPVC criteria.
Section 5.4.1.3 of the SDAA does not address susceptibility of the NPM to develop DWO conditions below 20% power for the secondary side. Section 5.4 indicates that DWO may occur on the secondary side at low power levels.
Given that the secondary side may be susceptible to DWO at low power levels, how is the COL item able to ensure that the COL applicant can demonstrate that the secondary side controller can ensure the reactor power response is stable for the NPM below 20% power? Given that the NPM may be susceptible to DWO at low power levels, provide markups for the FSAR that that describe how the COL item is able to ensure that the COL applicant can demonstrate that the secondary side controller can ensure the reactor power response is stable for the NPM below 20% power or provide markups that update the COL item to include the development of a NuScale Nonproprietary NuScale Nonproprietary
methodology that can demonstrate that the secondary side controller can ensure that the secondary side will remain stable at power levels below 20%. Also, please confirm if the COL item is intended to address DWO and primary side stability.
Additionally, given that the secondary side may experience DWO as stated in 5.4.1.3 at low power levels, describe how the secondary side controllers are designed to function at low power levels where they may be experiencing stable conditions or oscillatory conditions (coherent/incoherent oscillations) in the FSAR markups for 7.0.4.5.
Response
The original response was posted on October 26, 2023. Following an in-person meeting at the NuScale offices in Rockville, MD, on April 18, 2024, NuScale agreed to revise the response to provide supplemental information. The original response below is unchanged. The supplemental information is added to the end of the response, starting with the section labeled Supplemental Response Following April 18, 2024 Meeting. The revised response, including the supplemental information and a Final Safety Analysis Report (FSAR) markup, was posted on May 29, 2024.
The NRC provided feedback on July 29, 2024 stating: ((2(a),(c) The COL Item 7.0-1 for the US460 standard design approval application (SDAA) originated from COL Item 7.0-1 for the US600 design certification application (DCA). During the NRC review and approval of the DCA, COL Item 7.0-1 was added by NuScale in response to request for additional information (RAI) 9218 question 01-61 (submitted by NuScale letter RAIO-0219-58644, dated February 12, 2018) and discussed further in the NuScale response to RAI 9575 question 15.09-9 (submitted by NuScale letter RAIO-1018-62120, dated October 11, 2018). The purpose of adding COL Item 7.0-1 to the DCA was to defer analyses associated with the control systems because detailed design information on the control systems was not available during the DCA. The RAI 9218 question 01-61 indicated a gap in that effects of control system NuScale Nonproprietary NuScale Nonproprietary
operation had not been specifically evaluated for the potential to cause instabilities. The concern was the effect of such instabilities, if any, in terms of their consequences to the fuel. The COL Item 7.0-1 was not specifically added to the DCA to address density wave oscillations (DWO). NuScale included COL Item 7.0-1 in Revision 0 of the SDAA for the same reason it was included in the DCA: to defer analyses associated with the control systems because detailed design information on the control systems was not available. The COL Item 7.0-1 was not included in Revision 0 of the SDAA to specifically address DWO. NuScale believes that COL Item 7.0-1 is no longer necessary and proposes to delete COL Item 7.0-1 from the SDAA. The justification for deletion of COL Item 7.0-1 ((
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The Evaluation Methodology for Stability Analysis of the NuScale Power Module, TR-0516-49417-P-A, Revision 1, identifies a regional exclusion approach to stability. Stability is demonstrated by calculation of a decay ratio, with an acceptance criterion of less than or equal to 0.8, when single phase flow is maintained in the riser. In addition, module protection system (MPS) actuation is demonstrated to occur before riser subcooling is lost. The final safety analysis report (FSAR) for the SDAA provides the stability analyses performed in accordance with TR-0516-49417-P-A, Revision 1 in Section 15.9. These stability analyses demonstrate that changes in plant conditions do not result in instability or the MPS shuts down the reactor before any potential instability manifests itself as a divergent primary flow oscillation. ((
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(( }}2(a),(c) With the existing stability analyses in FSAR Section 15.9 and the supplemental evaluations to be performed to address the effects of control system malfunctions, there will be no analysis deferred to the COL scope. The SDAA analyses, plus the new discussion of the control system evaluations, satisfy the intended purpose of COL Item 7.0-1, which will no longer be needed. No markups are provided at this time; markups will be provided when the evaluation results are provided. Supplemental Response Following April 18, 2024 Meeting An in-person meeting between NuScale and the NRC was held at the NuScale offices in Rockville, MD, on April 18, 2024. During the meeting, NuScale presented results from two calculations: EC-153049, NPM-20 Analysis of SG Instability during Steady-State Operation, Revision 0, and EC-165065, Subchannel Analysis of Steam Generator Instability during Steady-State Operation, Revision 0. Following the meeting, EC-153049 and EC-165065 were placed in the electronic reading room (eRR). A summary of the calculations is provided as follows. (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary
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Overall, the two calculations demonstrate that the spectrum of existing Chapter 15 analyses bound the effects of an oscillation induced initiating event and that MPS effectively mitigates oscillations that have sufficient effect on the primary side conditions prior to the conditions challenging margin to specified acceptable fuel design limits (SAFDLs). Therefore, the calculations address the gap that effects of control system operation had not been specifically evaluated for the potential to cause instabilities that that could challenge specified acceptable fuel design limits and led to COL Item 7.0-1. With analysis no longer deferred, COL Item 7.0-1 can be deleted. This supplemental information, along with the information provided in the original response above, addresses the concerns identified in this audit question. Markups to FSAR Chapter 15 to incorporate relevant information from EC-153049 and EC-165065 are attached. Markups to FSAR Chapters 1 and 7 to delete COL Item 7.0-1 are also attached. Markups of the affected changes, as described in the response, are provided below: NuScale Nonproprietary NuScale Nonproprietary
NuScale Final Safety Analysis Report Stability NuScale US460 SDAA 15.9-2 Draft Revision 2 Reference 15.9-1, but with NPM parameters reflecting the US460 standard design, are presented in this section. Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 Although not described in the Reference 15.9-1 methodology, Section 15.9.5 presents evaluations of continuously-present, imposed oscillations on the secondary side through feedwater flow rate or main steam pressure. 15.9.2 Stability Analyses Several cases are analyzed over a wide range of power and primary system flow operating conditions and possible scenarios to demonstrate that stability is maintained during routine power operations in the NPM. 15.9.2.1 Stability Analysis for Power Operations This section analyzes power operations over a range of power level and flow conditions in the presence of a small perturbation in operating conditions as described in Section 8.1 of Reference 15.9-1. Only the limiting case is presented in detail in Section 15.9.2.1.1 through Section 15.9.2.1.4. The analyses show that the NPM is stable at power levels above 5 percent, with EOC conditions providing more damping (i.e. more stability) than BOC conditions. The flow stability condition that is the least stable occurs at 12.5 MW core power (5 percent of rated) with a BOC reactivity condition. The primary coolant flow response shows damped oscillations with a period of over a minute at this power level. The flow is less stable than the higher power cases, but with the low power level, there is no challenge to fuel limits. 15.9.2.1.1 Identification of Causes and Event Description The 12.5 MW power case at BOC is the least stable of the power operations considered. 15.9.2.1.2 Sequence of Events and Systems Operation No systems operations occur in response to the event. 15.9.2.1.3 Input Parameters and Initial Conditions Input parameters and initial conditions for the limiting case (5 percent power at BOC) are presented in Table 15.9-1.
NuScale Final Safety Analysis Report Stability NuScale US460 SDAA 15.9-8 Draft Revision 2 These results indicate that the NPM is stable during a postulated decrease in reactor coolant inventory. 15.9.4 Demonstration of Module Protection System Functions to Preclude Instability At rated power, the minimum loop time for the NPM is more than 60 seconds. The response delay for the MPS is no more than 8.0 seconds for setpoints that are pertinent to stability analysis and the reactor trip time is less than 2.5 seconds. The time from the first reactor trip setpoint being reached to the control rods being fully inserted is less than 11 seconds, which is significantly less than the minimum loop time for the NPM. Therefore, the MPS shuts down the reactor before any potential instability manifests itself as a divergent primary flow oscillation. Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 15.9.5 Secondary Side Oscillations Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 Section 15.9.2 and Section 15.9.3 demonstrate a perturbation in the secondary side does not result in instability in the primary side. In this section, evaluations are performed of continuously-present, imposed oscillations on the secondary side. The oscillations are not predicted to occur, but are instead artificially imposed as a boundary condition. The oscillations are continuously-present in that the oscillating boundary condition is applied throughout the evaluation regardless of the transient progression. Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 Continuously-present, imposed oscillations on the secondary side are evaluated, with consideration of the MPS response, using the NRELAP5 code. The impact of secondary side oscillations that are postulated to occur during steady-state operation are evaluated by imposing oscillations as a sine wave boundary condition. Cases are analyzed based on combinations of variations in the following parameters: Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 Secondary parameter with imposed oscillation is either steam pressure or feedwater flow, with the oscillation modeled to initially induce a cooling response from either initially decreased steam pressure or initially increased feedwater flow Oscillation amplitudes ranging from 5 percent to 50 percent Oscillation periods ranging from 5 seconds to 600 seconds Initial reactor power levels ranging from 25 percent to 100 percent Reactivity feedback response associated with BOC and EOC Automatic control rod bank response to primary coolant temperature disabled (all analyzed power levels) and enabled (100 percent power only) Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 The results show that main steam pressure changes have a larger impact on SG power than feedwater flow changes of the same relative magnitude. The larger
NuScale Final Safety Analysis Report Stability NuScale US460 SDAA 15.9-9 Draft Revision 2 change in SG power drives lower moderator temperature and a larger core power response. Cases starting from high initial power (i.e., 100 percent power) are most limiting for maximum power. The results show that high frequency (short period) oscillations in the secondary side do not translate well to the primary side. Secondary side oscillations with periods that are on the order of the primary side loop transit time (which varies with power level) can result in larger effects on primary side conditions. The range of oscillation periods considered is adequate to demonstrate that the most limiting response falls inside the range, considering the different primary loop transient times as a function of power. Longer oscillation periods, with sufficient magnitude, result in reactor trip without a full oscillation cycle, such that they resemble monotonic parameter changes similar to the increase in feedwater flow or increase in steam flow events in Section 15.1.2 and Section 15.1.3, respectively. The range of oscillation amplitudes considered is adequate to demonstrate that the most limiting response falls inside the range; increasing amplitude beyond a certain point results in earlier reactor trips with less limiting power responses. When automatic control rod bank response is disabled, EOC cases are more limiting because of reactivity coefficient feedback effects. When automatic control rod bank response is enabled, BOC cases are more limiting because the rod movement caused by the decreased moderator temperature drives a higher power response. Rod movement in response to a decreasing moderator temperature is modelled in the same manner as for increase in heat removal by the secondary system events in Section 15.1 (i.e., events with a cooldown of the RCS). Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 The cooldown of the RCS also causes increased density of the coolant in the downcomer and this increase in density can affect the power level detection by the excore neutron detectors. In order to account for this effect, the high power and high power rate trips are adjusted to account for downcomer coolant temperature. Although the maximum power of some cases is greater than the high power analytical limit in Table 15.0-7, the high power trip does not occur in the cases due to the adjustment of the analytical limit to account for downcomer coolant temperature. The results demonstrate that limiting transient conditions are mitigated by one or more of the following MPS signals: Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 high power rate high pressurizer pressure low main steam superheat high main steam superheat low main steam pressure low pressurizer pressure high RCS average temperature Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 A subset of the cases evaluated in NRELAP5 are further evaluated in VIPRE-01 to determine minimum critical heat flux ratio (MCHFR). The cases are selected from those with the most limiting power responses in NRELAP5 as well as from a range of other combinations of conditions to ensure the limiting MCHFR is captured. The
NuScale Final Safety Analysis Report Stability NuScale US460 SDAA 15.9-10 Draft Revision 2 analysis inputs and sequence of events for the limiting MCHFR case are provided in Table 15.9-13 and Table 15.9-14, respectively. The results for the limiting MCHFR case are shown in Table 15.9-15, and in Figure 15.9-11 through Figure 15.9-17. The limiting MCHFR case also has the highest peak power of the NRELAP5 cases. The VIPRE-01 results confirm the NRELAP5 findings above that the most limiting cases are those initiated from 100 percent power with oscillations imposed on main steam pressure and bank withdrawal enabled. The limiting MCHFR is associated with an oscillation-induced cooldown and subsequent rod withdrawal. Comparing the results and this limiting MCHFR to those from cooldown events in FSAR Section 15.1 and rod withdrawal events in FSAR Section 15.4 shows that the results are similar. The MCHFR for the limiting oscillation case is not more limiting than cooldown events analyzed in Section 15.1 and is not more limiting than rod withdrawal events analyzed in Section 15.4. Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 Overall, the evaluation demonstrates that the spectrum of analyses in Section 15.1 and Section 15.4 bound the effects of an oscillation-induced initiating event. The MPS effectively mitigates oscillations that have sufficient effect on the primary side conditions prior to the conditions challenging margin to specified acceptable fuel design limits. 15.9.6 Conclusions There are two main aspects of the stability methodology. The first is the use of a regional exclusion as the stability solution type and the rationale for its selection. The second aspect is the demonstration that the NPM maintains stability within the region of operation allowed by the MPS. The NPM returns to the original oscillation-free condition after steady state conditions are perturbed. Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 Operational events do not result in unstable NPM behavior or are terminated by the MPS prior to exceeding specified acceptable fuel design limits. Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 The radiological consequences of events in Section 15.9.3 and Section 15.9.5 are bounded by the design-basis accident analyses presented in Section 15.0.3. 15.9.7 References 15.9-1 NuScale Power, LLC, Evaluation Methodology for Stability Analysis of the NuScale Power Module, TR-0516-49417-P-A, Revision 1.
NuScale Final Safety Analysis Report Stability NuScale US460 SDAA 15.9-23 Draft Revision 2 Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 Table 15.9-13: Oscillation Spectrum Limiting Case - Inputs Parameter Input Value Initial power 100% Time in cycle BOC Automatic control rod bank response Enabled Secondary side parameter with imposed oscillation Main steam pressure Imposed oscillation amplitude +/-35% Imposed oscillation period 70 seconds Pressurizer pressure 2000 psia RCS flow rate Low end of range in Table 15.0-6 RCS average temperature 540°F SG pressure 475 psia Feedwater temperature 250°F Pressurizer level 60%
NuScale Final Safety Analysis Report Stability NuScale US460 SDAA 15.9-24 Draft Revision 2 Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 Table 15.9-14: Oscillation Spectrum Limiting Case - Sequence of Events Event Time [s] Steady-state operation 0 Main steam pressure oscillation imposed (beginning with a decrease in pressure) 30 Main steam pressure reaches local minimum (one quarter of one period) 47.5 Main steam pressure returns to original value (one half of one period) 65 Control rod insertion begins (from high power rate signal) 65 Peak reactor power occurs (122 percent) 65 Limiting MCHFR occurs 65 Decay heat removal system actuation, secondary system isolation, and pressurizer heater trip (from high pressurizer pressure signal) 67
NuScale Final Safety Analysis Report Stability NuScale US460 SDAA 15.9-25 Draft Revision 2 Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 Table 15.9-15: Oscillation Spectrum Limiting Case - Results Acceptance Criteria Limit Analysis Value MCHFR 1.45 1.62
NuScale Final Safety Analysis Report Stability NuScale US460 SDAA 15.9-36 Draft Revision 2 Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 Figure 15.9-11: Turbine Inlet Pressure for Limiting Oscillation Case 300 350 400 450 500 550 600 650 0 500 1000 1500 2000 2500 Turbine Inlet Pressure (psia) Time (sec)
NuScale Final Safety Analysis Report Stability NuScale US460 SDAA 15.9-37 Draft Revision 2 Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 Figure 15.9-12: Reactor Power for Limiting Oscillation Case 0 20 40 60 80 100 120 140 0 500 1000 1500 2000 2500 Core Power (%) Time (sec)
NuScale Final Safety Analysis Report Stability NuScale US460 SDAA 15.9-38 Draft Revision 2 Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 Figure 15.9-13: Reactor Coolant System Flow for Limiting Oscillation Case 0 200 400 600 800 1000 1200 1400 1600 0 500 1000 1500 2000 2500 RCS Flow (lbm/s) Time (sec)
NuScale Final Safety Analysis Report Stability NuScale US460 SDAA 15.9-39 Draft Revision 2 Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 Figure 15.9-14: Reactor Coolant System Average Temperature for Limiting Oscillation Case 525 530 535 540 545 550 555 0 500 1000 1500 2000 2500 RCS Average Temperature (F) Time (sec)
NuScale Final Safety Analysis Report Stability NuScale US460 SDAA 15.9-40 Draft Revision 2 Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 Figure 15.9-15: Reactor Coolant System Pressure for Limiting Oscillation Case 1900 1950 2000 2050 2100 2150 0 500 1000 1500 2000 2500 RPV Lower Plenum Pressure (psia) Time (sec)
NuScale Final Safety Analysis Report Stability NuScale US460 SDAA 15.9-41 Draft Revision 2 Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 Figure 15.9-16: Steam Generator Pressure for Limiting Oscillation Case (Typical of one Steam Generator) 200 400 600 800 1000 1200 1400 1600 0 500 1000 1500 2000 2500 Steam Pressure 1 (psia) Time (sec)
NuScale Final Safety Analysis Report Stability NuScale US460 SDAA 15.9-42 Draft Revision 2 Audit Question A-15-2, Audit Question A-15-4, Audit Question A-15-5 Figure 15.9-17: Critical Heat Flux Ratio for Limiting Oscillation Case
NuScale Final Safety Analysis Report Interfaces with Standard Design NuScale US460 SDAA 1.8-9 Draft Revision 2 Audit Question A-15-2 COL Item 7.0-1: Not used.An applicant that references the NuScale Power Plant US460 standard design will demonstrate the stability of the NuScale Power Module during normal and power maneuvering operations for closed-loop module control system subsystems that use reactor power as a control input. 7.0 COL Item 7.2-1: An applicant that references the NuScale Power Plant US460 standard design will implement the life cycle processes for the operation phase for the instrumentation and controls systems, as defined in IEEE Std 1074-2006 and IEEE Std 1012-2004. 7.2 COL Item 7.2-2: An applicant that references the NuScale Power Plant US460 standard design will implement the life cycle processes for the maintenance phase for the instrumentation and controls systems, as defined in IEEE Std 1074-2006 and IEEE Std 1012-2004. 7.2 COL Item 7.2-3: An applicant that references the NuScale Power Plant US460 standard design will implement the life cycle processes for the retirement phase for the instrumentation and controls systems, as defined in Institute of IEEE Std 1074-2006 and IEEE Std 1012-2004. The Digital I&C Software Configuration Management Plan provides guidance for the retirement and removal of a software product from use. 7.2 COL Item 9.1-1: An applicant that references the NuScale Power Plant US460 standard design will develop plant programs and procedures for safe operations during handling and storage of new and spent fuel assemblies, including criticality control. 9.1 COL Item 9.1-2: An applicant that references the NuScale Power Plant US460 standard design will provide the design of the spent fuel pool storage racks, including the structural dynamic and stress analyses, thermal hydraulic cooling analyses, criticality safety analysis, and material compatibility evaluation. 9.1 COL Item 9.1-3: An applicant that references the NuScale Power Plant US460 standard design will provide the periodic testing plan for fuel handling equipment. 9.1 COL Item 9.1-4: An applicant that references the NuScale Power Plant US460 standard design will describe the process for handling and receipt of critical loads including NPMs. 9.1 COL Item 9.1-5: An applicant that references the NuScale Power Plant US460 standard design will provide a description of the program governing heavy loads handling. The program should address
- operating and maintenance procedures.
- inspection and test plans.
- personnel qualification and operator training.
- detailed description of the safe load paths for movement of heavy loads.
9.1 COL Item 9.3-1: An applicant that references the NuScale Power Plant US460 standard design will submit a leakage control program for systems outside containment that contain (or might contain) accident source term radioactive materials following an accident. The leakage control program will include an initial test program, a schedule for re-testing these systems, and the actions to be taken for minimizing leakage from such systems to as low as practical. 9.3 COL Item 9.5-1: An applicant that references the NuScale Power Plant US460 standard design will provide a description of the offsite communication system, how that system interfaces with the onsite communications system, as well as how continuous communications capability is maintained to ensure effective command and control with onsite and offsite resources during both normal and emergency situations. 9.5 COL Item 10.3-1: An applicant that references the NuScale Power Plant US460 standard design will provide a site-specific Secondary Water Chemistry Control Program based on the latest revision of the Electric Power Research Institute Pressurized Water Reactor Secondary Water Chemistry Guidelines and Nuclear Energy Institute 97-06 at the time of the application. 10.3 Table 1.8-1: Combined License Information Items (Continued) Item No. Description of COL Information Item Section
NuScale Final Safety Analysis Report Instrumentation and Controls - Introduction and Overview NuScale US460 SDAA 7.0-15 Draft Revision 2 system HSIs are necessary. There are two boundaries between MCS and MPS: the fiber-optic isolated portion and the HWM boundary. The MCS has a direct, bi-directional interface with the PCS. The network interface devices for the MCS domain controller/historian provide the interface between the human-machine interface network layer and the control network layer. The MCS uses logic processing in the cases where redundant input or output channels are used. Some logic supports the redundant-channel architecture used by the MPS, while other logic directly supports the process systems. The logic processing of multiple channels can include two, three, or four input signals. Audit Question A-15-2 COL Item 7.0-1: Not used.An applicant that references the NuScale Power Plant US460 standard design will demonstrate the stability of the NuScale Power Module during normal and power maneuvering operations for closed-loop module control system subsystems that use reactor power as a control input. The potential for closed loop MCS subsystems that use reactor power as a control input to cause instabilities is addressed by evaluations in Section 15.9.5. The results in Section 15.9.5 demonstrate that MPS effectively mitigates potential oscillations that have sufficient effect on the primary side conditions prior to the conditions challenging margin to specified acceptable fuel design limits. Normal operation and power maneuvering control functions are provided by the following MCS functions for each NPM: turbine trip, throttle and governor valve control turbine bypass valve control feedwater pump speed control feedwater regulating valve control reactor coolant system boron concentration (chemical shim) control control rod drive system control pressurizer pressure control pressurizer level control The control inputs and functions for each during normal power operation are described below. Turbine Trip, Throttle and Governor Valve Control The turbine trip, throttle, and governor controls rely on the following control inputs: main turbine control system package sensors (case temperatures, drain valve position, eccentricity, speed sensing, shaft axial position, journal bearing displacement, journal bearing temperature and other sensors) demand power level (main turbine generator load or reactor power) from MCS and main turbine control system}}